ML20058A493
| ML20058A493 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 10/15/1990 |
| From: | Long R GENERAL PUBLIC UTILITIES CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 4410-90-L-0075, C000-90-109975, NUDOCS 9010260256 | |
| Download: ML20058A493 (64) | |
Text
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i GPU Nuoleet Corporet6en
?. !J Nuclear
- ,omers=
Middletown, Pennsylvania 17067 0191 717 944 7621 TELEX 84 2386 T1[N8N0 October 15, 1990 4410-90-L-0075 C000-90-1099
- US Nuclear Regulatory Commission
. Washington, DC 20555 Attention: Document Control Desk Three Mile Island Nuclear Station, Unit 2 (TMI-2)
Operating License No. DPR-73 Docket No. 50-320 Post-Defueling Monitored Storage Safety Analysis Report Amendment 0
Dear Sirs:
Attached is Amendment 8 to the Post-Defueling Monitored Storage (PDMS) Safety Analysis Report (SAR). _This amendment provides changed pages to the regulatory conformance section of the PDMS SAR and includes a request for an exemption-
- from the fracture toughness and material surveillance program requirements of 10 CFR 50.60. This request is submitted in accordance with the provisions of 10CFR50.12(a)forspecificexemptionswiththejustificationincludedinthe SAR text.
Sincerely, k00020 R. L. Long PD D
P PNV Director, Corpor Services /
Director, THI-2 EDS/mkk cc:
T. T. Martin - Regional Administrator, Region l' M. T. Masnik - Project Manager, TMI-2, PDNP Directorate L.- H. ~ Thonus' - Project Manager, TMI Site
.F.;I.-Young - Senior Resident Inspector, THI
..X 10i GPU Nuclear Corporation is a subsidiary of the General Public Utilities Corporation f
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CHAPTER 3
.g 3
DESIGN CRITERI A - STRUCTURES. ' SYSTEMS AND COMP (44.a 5 :
- H TABLE OF CONTENTS
-SECTION:
TITLE.
- PAGE L3.1 REGULATORY CONFORMANCE 3.1-1
)
f 3.l.1 CONFORMANCE WITH 10 CFR Part 50 3.1-1 3.1.2-GENERAL DESIGN CRITERIA
-3.1-24 3.2 CLASSIFICATION OF STRUCTURES, SYSTEMS, AND COMPONENTS 3.2-1 3'.2.1-SEISMIC CLASSIFICATION
-3.2-1 3.2.2-SYSTEM QUALITY GROUP CLASSIFICATION
- 3.2-3 l
3.3' WIND AND TORNADO LOADINGS 3.3-1 3.3.1 WIND LOADINGS 3.3-1 3.3.2 TORNADO LOADINGS 3.3-1 a
. 3.4 WATERLEVEL(FLOOD)' DESIGN 3.4-1 3.4.1 FLOOD ELEVATION 3.4-1 L
3.4.2 PHENOMENA CONSIDERED IN DESIGN LOAD CALCULATIONS 3.4-1 l
t 3.4.3 FLOOD FORCE APPLICATION 3.4-1 3.4.4 FL000 PROTEC110N 3.4-1 "3.5 MISSILE PROTECTION CRITERIA-3.5 I 3.5.1
. MISSILE LOADINGS AND BARRIERS 3.5-1 3 '. 5. 2 MISSILE SELECTION 3.5-1 3.5.3 SELECTED MISSILES 3.5-1:
i
-3.5.4 BARRIER DESIGN PROCEDURES 3.5-2
'3.5.5
-MISSILE BARRIER FEATURES 3.5-3 5
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. DESIGN CRITERIA - STRUCTURES, SYSTEMS,=.AND COMP 0NENTS' l
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SECTION:
TITLE' PAGE
' 3.1; REGULATORY CONFORMANCE 3.1-1 3.1.11 CONFORMANCE WITH 10 CFR-Part 50 3.1-1 GENERAL PROVISIONS 3.1.1.1 10 CFR 50.1 - Basis, purpose, and procedures 3.1-1 applicable.
' 3 '.1.1. 2 10 CFR 50.2 - Definitions.
3.1-1 3.1.1.3 10 CFR 50.3 - Interpretations ~.
3.1-1 3.1.1.4
-10 CFR 50.4 - Communications.
3.1 1 3.1.1.5-10 CFR 50.7 - Employee protection.
3.1-1 3.1.1.6
'10 CFR 50.8 - Information collection 3.1-2 requirements: OMB approval.
REQUIREMENT 0F' LICENSE, EXCEPTIONS 3.1.1.7 10 CFR' 50' 10 - License required.
3.1-2
'3.1.1.8 10 CFR 50.11 - Exceptions and exemptions 3.1 2-from-licensing requirements.
3;1.1.9 ;
I10'CFR 50.12 - Specific exemptions.
3.1-2 3.1.1~.10 '
10 CFR 50.13-Attacks and destructive acts by
~3.1-2 enemies of the United States; and defense-act'ivities.
CLASSIFICATION AND DESCRIPTION'0F LICENSES 3.1.1.11. ~10 CFR 50.20 - Two classes of licenses.
3.1-2
~3.1 ~. l.12 -
10-CFR_50.21 - Class 104 licenses;.for medical 3.1-2 therapy and'research and development. facilities.
i i
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, 2 CHAPTER ~3 I
8
- TABLE OF CONTENTS (Cont'd) a' np ' 1SECTION jTITLE 1A_GE f
3.1.1.13:
10 CFR 50.22 - Class 103 licenses; for.
3.1-2~
(
commercial and. industrial facilities.
3.1.1.14-10 CFR 50.23 - Construction permits.
3.1 ?
APPLICATIONS FOR LICENSES, FORM, CONTENTS, INELIGIBILITY Z
OF CERTAIN APPLICANTS
.i 13.1.1.15-10LCFR 50.30 Filing of applications for 3.1-3 1
~
licenses; oath or. affirmation.
3II.1.16, 10'CFR 50.31 - Combining applications.
3.1-3.
i 3.1.1.17 10 CFR 50.32 - Elimination of repetition.
3.1-3
]
3;1.1.18' 10 CFR 50.33 Contents of applications;-
3.1-3' t
general information.
3.1.1.19 10 CFR 50.33a - Information requested by 3.1-3
_the Attorney General.for antitrust-review.
.j t
3.1.1.20
- 10 CFR 50.34 - Contents of. applications; 3.1-3 q
L~
technical information, j
h l' '
3.1.1.21
-10 CFR 50.34a Design objectives for'-
3.1-9
[
equipment to control releases of radioactive-q
. material in effluents-nuclear power reactors.
l 1
3.1.1.22 10 CFR 50.35:- Issuance of construction permits.
3.1-9 3.1.'1. 23.
10 CFR 50.36 - Technical specifications.
3.1 '
3.1.1.24 10 CFR 50.36a - Technical specifications on 3.1-9' effluents from nuclear power reactors.
- 3.1.1.25
- ~10 CFR 50.36b - Environmental conditions.
-3.1-10:
3.1.1.26: '101CFR 50.37 Agreement limiting access to 3.1-10.:
iRestricted-Data
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CHAPTER-3 TABLE OF CONTENTS (Cont'd)
SECTION TITLE.
PAGE 3'1.1.27 10'CFR 50.38 - Ineligibility of certain 3.1-10 applicants.
3.1.1.28 10 CFR 50.39 - Public inspection of applications.
3.1-10 STANDARDS FOR LICENSES AND CONSTRUCTION. PERMITS 3.1.1.29 10 CFR 50.40 - Common standards.
3.1-10 3.1.1.30 10 CFR 50.41 - Additional standards for class 104 3.1-10 licenses.
3.1.1.31 10 CFR 50.42 - Additional standards for class 103 3.1-11 licenses.
3.1.1.32-10 CFR-50.43 - Additional standards and 3.1-11 provisions affecting class 103 licenses for commercial power.
3.1.1.33 10 CFR 50.44~~ Standards for combustible gas
.3.1-11 control system in light-water-cooled power reactors.
3.1.1.34 10~CFR 50.45 - Standards for construction permits.
3.1-11 3.1.1.35 10 CFR 50.46 - Acceptance criteria for emergency 3.1-11 core cooling systems for light water nuclear power reactors.
-~
3.1.1.36
'0 CFR 50.47 - Emergency plans..
3.1-11 3.1.1.37 10 CFR 50.48 - Fire protection.
3.1-12 3.1.1.38-10 CFR 50.49 - Environmental qualification of 3.1-12 electric equipment important'to safety for nuclear power plants.
ISSUANCE, LIMITATIONS, AND CONDITIONS OF LICENSES
.AND CONSTRUCTION PERMITS 3.1.1.39 10.CFR 50.50 - Issuance of licenses and 3.1-12 construction permits.
1 iii
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-CHAPTER 3.
- .' m
- .'TABLEOFCONTENTS<(Cont'd)
'SECTION-TITLE-PAGE 3.1'1.40 10 CFR 50.51 - Duration of license, renewal.
3.1 -
3.1.1.41' 10 CFR 50.52 - Combining' licenses.
3.1-12
'3.1.1.42 10 CFR 50.53
. Jurisdictional limitations.
3.1-12
-3.1.1.43 10 CFR 50.54 - Conditions of licenses.
3.1 3.1.1.44-10 CFR 50.55 - Conditions of construction permits.
3.1-16 3.1.1. 4 5 '-
10 CFR 50.55a - Codes and standards.
3.1-17 3.1.1.46 10 CFR 50.56 - Conversion of construction permit 3.1-18 to license; or amendment of license.
j 3.1.1.47 10 CFR 50.57 - Issuance of operating license.
3.1 3.1.1.48
'10 CFR 50.58 - Hearings and report of the Advisory 3.1-18 Committee on Reactor Safeguards.
3.1.1.49 10 CFR 50.59 -_ Changes, tests and-experiments.
3.1-18 3.1.1.50 10 CFR 50.60 - Acceptance criteria for fracture 3.1-19 prevention measures for light-water nuclear power reactors for normal operation.
3.1.1.51-10 CFR 50.61 - Fracture toughness requirements 3.1-19 for-protection against pressurized thermal sho'ck events.
3;1.1.52
-10 CFR 50.62 - Requirements for reduction-of risk 3.1-19 from anticipated transients without1 scram (ATWS) events for light-water-cooled nuclear power. plants.
3.1;1.53 -
10 CFR 50.63 - Loss of alternating current power 3.1-20 INSPECTION, RECORDS, REPORTS, NOTIFICATIONS 3.1.1.54L 10 CFR 50.64 - Limitations on the use of highly 3.1 enriched-uranium.(HEU)in'domesticnon-power. reactors-3.1.1.55-10 CFR 50.70 - Inspections.
3.1-20 iv
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g
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.c.-
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CHAPTER 3
-8 '
.TABLEOFCONTENTS;(Cont'd)
SECTION
' TITLE.
PAGE-
',, 3.1.1.56 10 CFR 50.71 - Maintenance of records, making of 3.1-20 reports.
3.1.1.57 10 CFR 50.72 - Immediate notification requirements 3.1-21 for operating nuclear power reactors.
3.1.1.58 10 CFR 50.73 - Licensee event report-system.-
3.1-21 W
3.1.1.59 10 CFR 50.75 - Reporting and Recordkeeping for 3.1-22 decommissioning planning n
-US/IAEA SAFEGUARDS AGREEMENT 3.1.1.60 10 CFR 50.78 - Installation information 3.1-22 and verification.
TRANSFERS 0F LICENSES-CREDITORS RIGHTS-SURRENDER OF LICENSES 3.1.1.61 10 CFR 50.80 - Transfer of licenses.
3.1-22 3.1.1.62
'10 CFR 50.81 - Creoitor regulat>ons.
3.1-22 3.1.1.63 10 CFR 50.82 - Applications for termination 3.1-22 of 1 Menses.
.AMENDi;ENT OF LICENSE'OR CONSTRUCTION PERMIT AT REQUEST OF HOLDER 3.1'.1.64 10 CFR-50;90 - Application for amendment ofi 3.1-22 license-or construction permit.
3.1.1.65 ~ 10 CFR 50.91 - Notice for public comment; 3.1-22 state consultation..
3'.1.1.66 10 CFR 50.92 - Issuance of amendment.
3.1-23 REVOCATION,: SUSPENSION, MODIFICATION, AMENDMENT.0F LICENSES-AND CONSTRUCTION PERMITS ~, EMERGENCY OPERATIONS'BY~THE COMMISSION
-3.1.1.67 10 CFR 50.100 - Revocation, suspension, 3.1-23 modification of licenses and construction permits-.for cause, y
_m._.
j
-., f ' -
CHAPTER'3
- "Y
. TABLE'0F-CONTENTS '(Cont'd)
SECTION TITLE PAGE M,
>3.1.1.68 10 CFR 50.101 - Retaking possession of special 3.1-23 nuclear material.
3.1.1.69-10 CFR 50.102 - Commission order for 3.1-23
. operation after revocation.
3.1.1'.70
'10 CFR 50.103.- Suspension and operation in 3.1-23 war or national-emergency..
BACKFITTING
'3.1.1.71 10 CFR 50.109:- Backfitting.
3.1-23 ENFORCEMENT 3.1.1.72 10 CFR 50.110 - Violations.
3.1-23 3.1.2 GENERAL DESIGN CRITERIA 3.1-24 3.1.2.1-Criterion 1 - Quality standards and records.
3.1-24 3.1.2.2:
Criterion 2 - Design bases for protection against
-3.1-25 natural phenomena.
'3.1.2;3-Criterion 3 - Fire protection.
3.1-25 3.1.2.4:1 Criterion 4-- Environmental and missile design 3.1-26 bases.
3.1.2.5;
. Criterion 5 - Sharing of structures,- systems, 3.1-26
-and components.
3.1!2.6-Criterion 101-Reactor design.
3.1-27
' 3.1.2.7-Criterion 11 --Reactor inherent protection.
3.'l-27
- +
3.1.2.8 Criterion ~12 - Suppression of reactor power 3.1-27 oscillations.
n.
!3.1.2.94 Criterion?l3 --Instrumentation and control.-
3.1-27 3.1~2.10-Criterion 14 - Reactor coolant pressure boundary.
3.1-28 vi m
y(
,-g_"
- CHAPTER 3-4.
TABLE OF' CONTENTS' (Cont'.d)
~SECTION;
' TITLE; PAGE-l Criterion 15:- Rea'ctor, coolant system design.
3.1-28 1
' 3.1.2.11 l3.1.2.12_. Criterion 16 - Containment design.
3.1-28
'3.1.2.13-Criterion 17 - Electric power systems.
3.1-29
-[
- 3.1.2.14 Criterion 18 - Inspection and testing of electric 3.1-29 power systems.
3.1.2.15 Criterion 19 - Control room.
3.1-30 3.1.2.16 Criterion 20 - Protection system functions.
3.1-30 3'.1. 2.17 - ' Criterion 21 - Protection system reliability and 3.1-30 testability.
'3.1.2.18 Criterion Protection system independence.
3.1-31 3.1.2.19 Criterion 23 - Protection system failure modes.
3.1-31 3.1.2.20 ' Criterion 24 - Separation of protection and 3.1-31
, control systems.
3.1.2.21 Criterion 25 - Protection system requirements 3.1-32 for reactivity control malfunctions.
4 3.'1.2. 22 -
Criterion 26 -. Reactivity control system 3.1-32 redundancy and capability.
3.1.2.23- ' Criterion 27 - Combined reactivity control 3.1-32 systems capability..
3.l.2.24 Criterion 28 - Reactivity limits.
3.1-35 3.1.2.25' Criterion 29 - Protection against anticipated 3.1-33 operational occurrences.
3.1.2.26 - Criterion 30 - Quality of reactor coolant 3.1-33 pressure boundary.
3.1.2.27
' Criterion 31 - Fracture prevention of reactor 3.1-33; coolant pressure boundary ~.
vii
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- CHAPTER 3, TABLEOFCONTENTSf(Cont'd) t SECTION TITLE
.P!GE
}[
~
3;1.2.28' Criterion 32 - Inspection of reactor coolant-3.1-34 pressure boundary.
J3.1.2.29L Criterion 33 - Reactor coolant makeup.
3.1 3.1.2.30-Criterion 34 - Residual heat removal.
'3.1-34:
3.1.2.31 Criterion'35 - Emergency core cooling.
.3.1-35
~3.1.2.32 Criterion 36 - Inspection of emergency core 3.1-35' cooling system.
y 3.1.2.33 Criterion 37 - Testing of emergency core cooling 3.1-35 system.
3.1.2.34 Criterion 38 - Containment heat removal.
3.1-36 3.1.2.35 Criterion 39 - Inspection of containment heat 3.1-36 removal system.
3.1.2.36 Criterion 40 - Testing of containment heat 3.1-36
removal system.
3.1.2.37 Criterion 41 - Containment' atmosphere cleanup.
3.1-37 3.1.2.'38 Criterion 42 - Inspection of containment 3.1-37 atmosphere cleanup systems.
3.1.2.39-Criterion 43 - Testing of containment atmosphere 3.'l-37 cleanup systems.
x
- 3.1'.'2. 40 Criterion 44 - Cooling water.
-3.1-38:
3.1.2.'41 Criterion 45 - Inspection of. cooling water system.
3.1-38 y
3.1.2.42 Criterion 46 - Testing of cooling water system.
3.1-38:
4
.3.1.2.43 Criterion 50'- Containment design basis.
3.1-39 3.1.2.44 Criterion'51 - Fracture prevention of-3.1,
containment pressure boundary.
3.1;2.45-Criterion 52 - Capability for-containment 3.1-40 leakage rate testing.
viii 4
1 NQ~n
- - - - _ - - - -. - - -. - - - -, -. _ _. - _ _ - _ - - - _ - _ - - - - - _ - - _ - - _ _. _ - - _. _ - - _ - _ - _ _ -. _ _ ~ - - -. - _ -. _ - _ - - - - _ _ - - _ - - - _ -. _ - - - - _. - _ _ - _ -
n i
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CHAPTER 3 TABLEOFCONTENTS.(Cont'd)
S SECTION TITLE-PAGE
-3.1.2.46
-Criterion 53 - Provisions for containment 3.1-40 testing and inspection.
'3.1.2.47: ' Criterion 54 - Piping systems penetrating 3.1-41 containment.
3.1.2.48 Criterion 55 - Reactor coolant pressure boundary 3.1-41 penetrating containment.
3.1.2.49 Criterion 56 - Primary containment isolation.
3.1-42
~ 3.-l. 2. 50 Criterion 57 - Closed system isolation valves.
3.1-43 3.1.2.51 Criterion 60 - Control of releases of 3.1-43 radioactive materials-to the environment.
3.1.2.52 Criterion 61 - Fuel storage and handling and 3.1-43 radioactivity control.
3.1.2.53 Criterion 62 - Prevention of criticality in fuel 3.1 -storage and handling.-
-3.1.2.54 Criterion 63 - Monitoring fuel and waste storage.
3.1-44
-3.1.2.55 Criterion 64 - Monitoring radioactivity releases.
3.1-44 2
ix
1 I
IE Y CHAPTER 3
' DESIGN CRITERIA - STRUCTURES, SYSTEMS, AND COMPONENTS
- 3;1' REGULATORY CONFORMANCE 3.1.1
.-CONFORMANCE WITH 10 CFR Part 50 Three Mile Island Nuclear Station Unit 2 was originally designed to confonn to the-Regulations' of ~ 10 CFR Part 50,_ including the General Design Criteria of Appendix A.
On March 28, 1979,.the unit experienced an accident which severely damaged the reactor core.. Subsequently, the core was removed and shipped off-site. The removal of-the reactor core and the revision of the license to a non-operating license have changed the function of the facility from an operating nuclear power:
plant to_ one 'of management and maintenance. - These characteristics substantially alter the applicability. and the requirement for Three Mile Island Unit 2 conformance with the regulations of 10 CFR Part 50. The. degree and manner of addressing _those regulations which are applicable to Three Mile Island Unit.2 during Post Defueling Monitored Storage are described in the following sections.
In addition, based on this evaluation, a request for relief from the-inservice inspection requirements of 10 CFR 50.55a and an exemption from the requirements of 10 CFR 50.60 are also provided. The-regulations which have been reviewed are those which were revised-as of January 1, 1990.
3.1.1.1 10_CFR 50.1 --Basis, purpose, and procedures applicable.
Article 50.1 describes the basis, parpose, and procedures of 10 CFR Part 50.
No exceptions are taken to the provis'.ons of this article.
3.1.1.2
.10 CFR 50.2 - Definitions.
-Article 50.2 provides definitiens of terms used throughout 10 CFR 50.
No exceptions are taken to the provisions of this article.
3.1.1.3 10 CFR 50.3 - Interpretations.
- Article' 50.3 delegates the responsibility for interpretations of~ 10 CFR 50 to the LGeneral Counsel. No exceptions'are taken to the provisions of_this article.
3.1.1.4 10 CFR 50.4 --Communications.
- Article 50.4 describes communications requirements for nuclear power plants.
No exceptions are taken to_the provisions of this article.
3.1.1=.5
_10 CFR 50.7 - Employee protection.
ArticleL50.7 describes provisions relating to the protection of employees at
. nuclear facilities.
No exceptions are taken to the provisions of this article.
3.1-1 AMENDMENT 8 - OCTOBER 1990
56.34(b)(2)(ii)-
3
-Paragraph 50.34(b)(2)(ii) requires the discussion of a number of items (e.g.,
ventilation systems, control systems, waste handling) for facilities other than nuclear reactors. Although TMI-2.was originally licensed as a nuclear reactor,,
during PDMS the. facility will.not function.as a nuclear reactor. Due to the unique condition cf TMI-2 during PDMS, the provisions described in this paragraph fmore~ accurately portray requirements for THI-2 than does paragraph 50.34(b)(2)(i)..
The intent of the provisions of this paragraph, as they relate to TMI-2 has been a
addressed in this document with the additional consideration:that TMI-2 was originally licensed as a nuclear. power reactor.
50.34(b)(3)
Paragraph 50.34(b)(3) requ res that the kinds of radioactive materials produced; and the means for controll ng and limiting effluents and exposures be described.
Although there will be no radicactive materials produced during PDMS as would normally occur at an opera:ing pwer reactor, there will be some radioactive waste generated.as well as the residual contamination that remains within the facility, s
Although the specific requirements of this paragraph cannot be complied with as a
written due to the unique condition of TMI-2 during PDMS, the intent of the l
provisions of this paragraph has been addressed by providing descriptions of the i
kinds of' radioactive materials which remain at the facility and the means-for j
controlling and limiting effluents and exposures to these materials to within the a
L limits of 10 CFR 20.
50.34(b)(4) l Paragraph 50.34(b)(4) requires a final analysis and evaluation of the structures,-
'J systems and components-which relate to-the protection of the public from the L
consequences of normal operation, transients, and accidents.- Although the R
. specifics _of these requirements as given in paragraph 50.34(b)(4) do no_t apply for PDMS,Lthe intent of these-requirements has been addressed for the. limited number
-of p.ostulated events and the insignificant risk to public health and safety.has j
been demonstrated.
]
L Paragraph 50.34(b)(4) also requires an analysis and-evaluation of ECCS cooling-L performance following postulated loss-of-coolant accidents in accordance.with the j
= requirements of 10 CFR 50.46. As 10 CFR 50.46 does not apply.to THI-2 in.its l
-current defueled condition, an analysis of ECCS cooling performance is not
'j
- provided, j
L 50.34(b)(5) l Paragraph 50.34(b)(5) requires'the description and evaluation of programs to resolve safety questions identified at the construction permit stage. These 1
L requirements were' addressed in the FSAR and do not apply to TMI-2 during PDMS.
l-a i
L 3.1-5 AMENDMENT 8 - OCTOBER 1990 l
ji 50.34(b)(6)('i)-
A Paragraph.50.34(b)(6)(i)irequires:information concerning<the applicant's organizational structure, allocations of responsibilities and authorities, and personnel _ qualifications requirements. Although the actual organization and responsibilities will;beLsubstantially different than that of a normally operating.
' power plant, the requirements of: this paragraph as they apply to TMI-2 during PDMS have been addressed in Section 10.5.-
l 50.34(b)(6)(ii)
.r Paragraph 50.34(b)(6)(ii) requires a description of the managerial and administrative controls used to satisfy applicable requirements of 10 CFR Part 50 Appendix B.
Appendix B establishes quality assurance requirements for activities affecting the-safety-related functions of those structures,-systems and components
-that. prevent or mitigate the consequences of postulated accidents that could cause undue risk to the health and safety of the public. During PDMS, TMI-2 has no structures, systems or components classified as safety-related and, therefore, the requirements of paragraph 50.34(b)(6)(ii)-and Appendix B do not apply to TMI-2 i
during PDMS. Due to the unique condition of THI-2 during PDMS, the specific requirements of this paragraph are not directly applicable; however, TMI-2 has addressed the intent of this paragraph by establishing a quality assurance program.
e similar to that described in Appendix B for activities such as radioactive waste 1
shipping and conformance with 10 CFR 20 requirements as well as for all activities 1
which are judged to be within the intent of this paragraph.
-l 50.34(b)(6)(iii)
! Paragraph.50.34(b)(6)(iii) requires information concerning plans for preoperational l
testing and initial operations. These' requirements do not' apply to TMI-2 during PDMS.
]
50.34(b)(6)(iv)
Paragraph 50.34(b)(6)(iv) requires information concerning' plans for conduct of normal operations, including maintenance, surveillance and' periodic testing of--
structures, systems and components.- During PDMS, operations'are limited to those l
activities related to monitoring and maintaining the facility in a stable..
}
condition.- Although the specific requirements of;this paragraph do not apply due 1
to the unique condition of TMI-2 during PDMS, the intent of these provisions has been addressed by providing information.concerning' activities appropriate during l
1 PDMS.-
50.34(b)(6)(v) j Paragraph 50.34(b)(6)(v). requires information concerning plans for coping with emergencies, which shall include the items specified in Appendix E.
Emergency
, planning requirements are based on the assumption of the potential necessity to notify the public of the existence of, or potential for significant 3.1-6 AMENDMENT 8 - OCTOBER 1990
~
4
9 'ff-site releases. Appendix-E recognizes that emergency planning needs are L.'
0Edifferent for, facilities that present less risk to the public., Due to the f
tnon-operatingLand defueled status of TMI-2 dur',g PDMS,-there is no potential fori any significant off-site radioactive-release.
Further.due to the existence of
- TMI-1 on'the'same site, emergency planning requirements for the site are dominated
~by-TMI-1 > Therefore, the-limited emergency planning _necessary_to accommodate the-existence'of THI-2 on the'same-site as_THI-1 has been incorporated in an integrated corporate emergency plan.
50.34(b)(6)(vi).
Paragraph 50.34(b)(6)(vi) requires information concerning proposed technical-specifications prepared in accordance with the requirements of Article 50.36.
Due to the-unique condition of TMI-2 during PDMS, the specific requirements _of Article 50.36 are.not applicable; however, the intent of this' article =has been addressed'by providing1 Technical Specifications in Chapter 9 which reflect the unique technical specification requirements for_PDMS.
-50.34(b)(6)(vii)
Paragraph 50.34(b)(6)(vii) requires information concerning the construction of multiunit power plant-sites. Thesa requirements are not applicable to TMI-2 during PDMS.
50.34(b)(7)
Paragraph 50.34(b)(7) requires the SAR to include the technical qualifications of.
the applicant to engage in the proposed activities in accordance'with the regulations in this chapter. The technical qualifications of GPU Nuclear, which are applicable to activities related to the unique PDMS conditions, are provided in Section-10.5 50.34(b)(8)
Paragraph 50.34(b)(8) requires the SAR to-include a description and plans for implementation of an operator requalification program. The operator requalification program-shall, as a minimum, meet:the requirements for. those programs contained in Appendix A of Part 55.'of this chapter. Due to the
-nonoperating and defueled status of TMI-2 during PDMS, the requirements for
-licensed reactor operator.= do not apply and consequently the requirements for operator 1requalification also do not apply.
-50.34(b)(9)
Paragraph 50.34(b)(9) requires a description of protection provided against, pressurized thermal shock events. Due to the non-operat'ing and'defueled status of.
TMI-2-during PDMS, the requirements of paragraph 50.34(b)(9) do not apply.
In addition, THI-2 was granted-an exemption.to 10 CFR 50.61'(Reference _3.1-1)~which acknowledged t. hat THI-2 need take no measures-to protect'against pressurized thermal: shock.
3.1-7 AMENDMENT 8 - OCTOBER 1990
^'
g 1 50134(c)?
u iParagrap_h_50.34(c) requires 'each application for a license to. operate a production or utilization facility to include a physical security plan. Due to the unique.
< condition of TMI-2 during PDMS, the specific. requirements off this; paragraph are not' s
' applicable; however,-the intent of the requirements has been_ addressed:in this SAR.
LThe security provisions necessary for TMI-2'have been provided by locating the unit inside-the same protected area as TMI Unit 1 and the provisions incorporated in the 3
TMI site security plan referenced in Section 10.2.
.l.
H j
50.34(d) 1 Paragraph 50.34(d) requires that each application for a license to operate a h
production'or utilization facility that is subject to Article 73.50, Article 73.55, l
L or Article 73.60.shall include a licensee safeguards contingency plan in accordance L
with the criteria set forth in Appendix C to 10 CFR Part 73.
The safeguards 1
L contingency provisions necessary for TMI-2 are provided by being located inside the
-same protected area as'TMI-1 and are incorporated.in the safeguards contingency plan for the TMI site.
See Section 10.2.
50.34(e)
. Paragraph 50.34(e) requires that each applicant for a license to operate a h
production or utilization facility who prepared a physical security plan, a 1
safeguards contingency plan, or a guard qualification'and training plan shall j
protect the plans and other related Safeguards-Informati)n against unauthorized L
disclosure in accordance with the requirements of 10 CFR 73.21 as appropriate. Due L
to the non-operating and defueled status of TMI-2 during PDMS and the location of l
.TMI-1 on the same. site,.overall security will be controlled by the site security-L plan.
All security activities' established in accordance with the regulations ~in 10 f
CFR Part 50 will be protected against unauthorized disclosure in accordance with 10 l
CFR 73.21.
50.34(f)-
Paragraph 50.34(f) establishes TMI-related requirements for a specific group of; plants. _TMI-2 is not included in'this group of plants; therefore, this paragraph' l
'does not. apply to TMI-2.
50.34(a)
. Paragraph 50.34(g) requires applicants.for operating licenses docketed after May 17, 1982,- to include SRP evaluations with their license applications. - As this 1
application is not requesting an operating license for THI-2, this paragraph does not apply to'THI-2..
3.1-8 AMENDMENT 8 - OCTOBER 1990 i
Ll*
3.1.1.21; :10 CFR 50.34a;-_ Design objectives for equipment-to control releases of D:
radioactive material in effluents-nuclear power reactors.
Article 50.346 establishes requirements-for radioactive effluent control descriptions in construction permit and operating;11 cense' applications. Due to the
. unique condition of.TMI-2 during PDMS, the specific requirements of this article are not applicable; however,'there will be limited radioactive effluents to the environment.during PDMS. _ Descriptions of the-equipment to monitor _ and control' those releases are provided consistent with the intent of this article.
3.1.1.22 10 CFR 50.35 - Issuance of construction permits.
Article 50.35 establishes requirements for the Commission with respect to the issuance of construction permits and defines the limitations of the construction permit.
No exceptions are taken to the provisions of this article.
3.1.1.23 10 CFR 50.36 - Technical specifications
' Article 50.36 establishes requirements for Technical Specifications. Due to the-unique condition of TMI-2 during PDMS, the specific requirements of this article are not applicable; however, the intent of these requirements has been addressed by prov.iding Technical Specifications which address those aspects of the facility necessary to manage and maintain the passive isolation of the contamination which remains inside the containment. The proposed Technical Specifications are included I
in Chapter 9.
3.1.1.24 10 CFR 50.36a - Technical specifications on effluents from nuclear power reactors.
50.36a(a)
Paragraph 50.36a(a) establishes requirements for effluents for operating reactors.
Although THI-2 is not an operating-reactor and the requirements of this paragraph cannot be complied with as written, the effluents during PDMS will be controlled and. limited to very low values. The intent of the provisions of this paragraph'is
-addressed.by providing effluent limits and the description of how these limits will
.be met-in Chapters 7 and 8.
[
- 50.36a(a)(1)
Paragraph l50.36a(a)(1) requires 1 that procedures be developed for the control of effluents and-that equipment installed in the radioactive waste system pursuant to 50.34(a):be maintained and used.
Procedures will be in place.for the control-'of -
effluents.during PDMS. The THI-2 equipment that will be used to process radioactive wastes during PDMS is described in-Section 7.2.3.
3.1-9 AMENDMENT 8 - OCTOBER 1990.
I' 56136a(a)(2)
- =..
-Paragraph:50.36a(a)(2) requires that each licensee. submit semi-annual reports on effluents and. prepare estimated public dose from those effluents. These requirements are applicable to THI-2 during PDMS.
50.36a(b)
Paragraph 50.36a(b) establishes guidelines for limiting radioactive effluents:and references 10 CFR 20.106 and 10 CFR 50 Appendix I as applicable in. limiting
-effluents. These requirements are applicable to THI-2 during PDMS.
3.1.1.25 10 CFR 50.36b - Environmental conditions.
Article 50.36b establishes that the NRC may specify conditions as part of the license to protect the environment.
No exceptions are taken to the provisions of-this article.
3.1.1.26 10 CFR 50.37_ - Agreement limiting access to Restricted Data.
Article 50.37 establishes-requirements for access to Restricted Data. No exceptions are taken to the provisions of this article.
3.1.1.27 10 CFR 50.38 - Ineligibility of certain applicants.
Article 50.38 establishes that certain persons are not eligible to apply.for or-obtain a license. No exceptions are taken to the provisions of this article.
3.1.1.28 10 CFR 50.39 - Public inspection of applications.
Article 50.39 states that applications and documents submitted to the Commission may be made available for public inspection. No exceptions are'ta'in to the
. provisions of'this article.
3.1.1.29 10 CFR 50.40 - Common standards.
- Article 50.40 establishes guidelines for the Commission in determining if a License will-be issued to an applicant.
No exceptions are taken to the provisions of this: article.-
3.1.1.30 10 CFR 50.41 - Additional standards for class 104 licenses.
Article 50.41 establishes additional standards for class 104 licenses for the Commission to use in determining if a license will be issued to an applicant. The class of license described in this article does not apply to TMI-2, 3.1-10 AMENDMENT 8 - OCTOBER 1990 l
m 1
( *3'.1'.1 ~.311 10lCFR50.42-'Additionalstandardsforclass103 licenses.
- c
. Article 50.42 establishes add t onal standards for class 103 licenses for the ii Commission to use in-determining if a license will be issued to an applicant.
No exceptions are;taken to the provisions of this article.
3.1.1.32 10 CFR 50.43 - Additional standards and provisions affecting class 103 licenses for commercial power.
Article 50.43 establishes additional standards and provisions for class 103 licenses. No exceptions are taken to the provisions of this article.
3.1.1.33 10 CFR 50.44 - Standards for combustible gas control system in light-water-cooled power reactors.
Article 50.44 establishes requirements for a combustible gas control system to be
-used in the event of a LOCA for each light water nuclear power reactor fueled with oxide pellets within cylindrical zircalloy' cladding. Due to the non-operating and defueled status of TMI-2 during PDMS, the TMI-2 reactor will not be fueled with oxide pellets within cylindrical zircalloy cladding. Thus, these requirements are not applicable to TMI-2, 3.1.1.34 10 CFR 50.45'- Standards for construction permits.
Article 50.45 establishes standards for the issuance of a construction permit. No exceptions are taken to the provisions of this article.
3.1.1.35-10 CFR 50.46 - Acceptance criteria for emergency-core cooling systems for light water nuclear power reactors.
. Article 50.46 defines acceptance criteria and requirements for emergency core cooling systems for light water nuclear power reactors fueled with oxide pellets wi_ thin cylindrical zircalloy cladding. Due to the defueled and non-operating status of THI-2 during'PDMS, the~TMI-2 reactor will not be fueled with uranium oxide pellets within cylindrical zircalloy cladding. Thus, the requirements of this article do not apply.
- 3.1.1.36 10 CFR 50'.47 - Emergency plans.
Article 50.47 establishes requirements for the content and criteria for acceptance of emergency plans.
Emergency-planning requirements are. based on the assumption of the potential necessity to notify the public of the existence of, or potential for
-significant;off-site releases. Appendix E recognizes that emergency planning needs are different for facilities that present less risk to the public.
Due to the
~non-operating and defueled' status of THI-2 during PDMS, there is no potential for any significant off-site radioactive' release. Due to the existence of TMI-1 on the same site, emergency planning.requirementsifor the site are dominated by TMI-1.
Therefore,' the limited emergency planning-necessary to accommodate the existence of THI-2.on the same site as TMI-1 has been incorporated in an integrated corporate emergency plan.
See~thediscussionofparagraph50.34(b)(6)(v).
3.1-11 AMENDMENT 8 - OCTOBER'1990
k j
'3'1.1.37
_10 CFR 50.48'- Fire protection.
L'
' Article 50.48_ establishes-requirements for fire protection 1for.an operating power plant. Due to the non-operating and defueled-status of TMI-2 during PDMS, the requirements-of this paragraph co not apply; however, the intent of this article
' has:been addressed by-providing fire protection consistent with the status of the facility.
- 3.1.1.38 10 CFR 50.49 - Environmental qualification of electric equipment important'to safety for nuclear power plants.
Article 50.49 establishes requirements for each holder of a license to operate a nuclear power plant to establish a program for the qualification of electric equipment important to safety.
Further, paragraph 50.49(b)defineselectric
~
equipment important to safety as tha', equipment relied upon to remain functional during design basis events to ensure: (i) the integrity of the reactor coolant pressure boundary, (ii) the capabil ty to shutdown the reactor and maintain it in a safe shutdown condition and (iii) tFe capability to prevent or mitigate the l
consequences of accidents that coulo result in potential off-site exposures comparable to the 10 CFR Part 100 guidelines. Due to the nonoperating and defueled status of TMI-2 during PDMS there is no electrical equipment which'is classified important to safety in accordance wita the three criteria established in paragraph 50.49(b).
Therefore the requirements of Article 50.49 do not apply to TMI-2 during PDMS.
In addition, the NRC granted THI-2 an extension to the compliance date for 10 CFR 50'49 (Reference 3.1-2) until six months prior to TMI-2's anticipated return
-to power.
Since this application eliminates the legal authority to operate the THI-2' facility _from the license, a subsequent license application would be necessary to resume-power operation. Therefore, during PDMS, no further action is required.
3.1.1.39 10.CFR 50.50 - Issuance of licenses and construction permits.
-Article 50.50 states that the Commission will issue a license or construction permit with such conditions and limitations as it deems appropriate.
No exceptions are_taken to the provisions of this article.
3.1.1.40 10 CFR 50.51 - Duration of license, renewal.
Article-50.51 establishes the durations of licenses issued by the Commission. No-exceptions are taken to the provisions of'this article.
3.1.1.41 10 CFR 50.52 - Combining licenses.
Article:50.52 establishes that the Commission may combine _ licensed activities in a single license.
No exceptions are taken to the provisions <>f this article.
3.1.1.42 10 CFR 50.53 - Jurisdictional limitations.
Article 50.53 establishes jurisdictional limitations on licenses.
No exceptions-are taken to the provisions of this article.
3.1-12 AMENDMENT 8 - OCTOBER 1990 l
d 9.!
'35.5.1.43 10 CFR'50.54' Conditions of licenses.
~
-1 Article 50.54 establishes.a series of cond'itions applicable.to holders of 'a
!1icense. Que to the non-operating.and defueled status of TMI-2 during PDMS, many 4
-of these requirement's do not apply. The. applicability of each paragraph of Article y
50.54 has been addressed =in the following review.
50.54(a)
Paragraph 50.54(a) requires that each nuclear power plant or_ fuel reprocessing.
plant licensee subject to the criteria of 10 CFR Part 50 Appendix-B implement a-quality assurance program pursuant to paragraph 50.34(b)(6)(ii).
Appendix B
' establishes quality assurance requirements for the safety-related functions of those structures, systems.and components that prevent or mitigate the consequences of postulated' accidents that could cause undue risk to the health and safety of the l
public. During PDMS, TMI-2_will not have any-structures, systems or components
, classified as safety-related and, therefore, the requirements.of paragraphs.-
50.54(a)(1), 50.34(b)(6)(ii) and Appendix B do'not apply to TMI-2.
However,.the intent of this article has been addressed by establishing and maintaining a quality assurance program similar to that described in Appendix B for TMI-2 activities.-
1 50.54(b) through 50.54(h) l Paragraphs 50.54(b) through 50.54(h) establish general limitations on licenses.- No R
- exceptions are taken to the provisions of these paragraphs.
50.54(i) through-50.54(m)
-Paragraphs 50.54(i)through50.54(m)establishrequirementsrelatedtoreactor operators and senior reactor operators. As discussed in License Amendment No. 30
.(Reference 3.1-3), these requirements'are specified for fueled reactors'.
As the
- TMI-2 reactor has been_defueled, the requirements of.these paragraphs do not apply to THI-2<during PDMS.
AlsoseeSection3.1.1.20regardingparagraph50.34(b)(8),
l L
'50.54(n)
Paragraph 50.54(n) states _ that "The licensee shall not, except as authorized pursuant.to a construction permit, make any alteration in the' facility constituting
' a_ change from;the technical specifications previously incorporated in a license or constructio_n permit pursuant to Article 50.36 of this part." No exceptions are-taken_to the provisions of the' article, i
I 3.1-13 AMENDMENT 8 - OCTOBER 1990 4
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,,+
- =
H 1
. 56.54(o) i
- Paragraph 50.54(o) states that4 primary reactor containments shall be_ subject to the
- requirements of 10.CFR 50. Appendix J.
The NRC granted T!il-2 an exemption (Reference 3.1-4) from the testing requirements of ApNndix J with the. exception that airlock door seals shall be tested in accordanca with Appendix J subsection III.D.2.b.iii._ This subsection requires airlock door seal testing during periods
, hen containment integrity is requirsd. As conteinment integrity is not required w
during PDMS this requirement is re longer applicable. However, as containment isolation will be required during PDMS, the proposed PDMS Technical Specifications-
-l
' do contain an appropriate Technical Specification for ensuring airlock door-operability.
j 50.54(p)
Paragraph 50.54(p) requires that a licensee prepare and maintain safeguard-contingency plan procedures and provides for revisions to those procedures. The safeguards contingency provisions necessary for THI-2 are provided by being _ located inside the same protected area as THI-1 and are incorporated in-the safeguards contingency. plan for the TMI site. See Section 10.2.
50.54(a)
~!
Paragraph 50.54(q) requires that a licensee shall follow and maintain emergency plans which meet the requirements of paragraph 50.47(b). This paragraph also defines requirements for revising those emergency plans. Due to the existence of TMI-1 on the samc~ site as THI-2, emergency planning requirements for the site are dominated by TMI-1.
Therefore, the limited emergency planning necessary to accommodate the existence of TMI-2 on the same site as TMI-1 has been incorporated l
in an -integrated corporaw ~ mergency plan, e
50.54(r)
- Paragraph 50.54(r) establishes requirements for test reactors. These requirements do not apply to THI-2.
u 50.54(s)
Paragraph 50.54(s) requires each licensee _who is authorized to possess and/or 1
operate _a nuclear power reactor to submit radiological emergency _ plansLof state and-local governmental entities to_the NRC. All radiological emergency planning provisions necessary for THI-2 have been incorporated in the THI site emergency.
planning process, including the provisions of paragraph 50.54(s).
I 50.54(t)
- Paragraph 50.54(t) establishes requirements _ for the development, revision, implementation and maintenance of the emergency preparedness program for nuclear-power reactors.
Emergency preparedness requirements applicable to TMI-2 are-incorporated in the emergency preparedness program established for the TMI-site.
See Section 10.3.
3.1-14 AMENDMENT 8 - OCTOBER 1990 l
1
I 75d.54(u)-
- Paragraph 50.54(u)~ requires each licensee 'to submit emergency, plans _ in' accordance i ^
with 10 CFR 50.47(b) and: Appendix E.
Article 50.47 establishes requirements for the~ content and criteria:for= acceptance of emergency plans.
Emergency planning W
7 requirements are-based.on the assumption of the potential necessity to notify the-1 public of the. existence of, or potential for significant off-site releases.
~ Appendix E recognizes that emergency planning needs are different for facilities that'present less risk to the public. Due to the non-operating and defueled status of TMI-2 during.PDMS there is no potential for any significant off-site radioactive
-release and due to the existence of TMI-1 on the.same site, emergency planning
- requirements for the site will be dominated by TMI-1. Therefore, the limited-emergency planning necessary to accommodate the existence of TMI-2 on.the same site as THI-1 has been incorporated in an integrated corporate emergency plan.
See
. Section3.1.1.20regardingparagraph50.34(b)(6)(v).
50.54(v)
Paragraph 50.54(v) requires that each licensee shall ensure that physical security, safeguards contingency and guard qualification and training plans and other related
- safeguards information are protected against unauthorized disclosure in accordance with the requirements of 10 CFR 73.21 as appropriate. To the extent that TMI-2 possesses the above. information during PDMS, it will be protected from unauthorind disclosure in accordance with 10 CFR 73.21.
See paragraphs 50.34(c), 50.34(d) and 50.34(e).
50.54(w)
Paragraph 50.54(w) requires that each electric utility licensed under this-part for a production or utilization facility of the type described in paragraph 50.21(b) or paragraph 50.22 shall by June 29, 1982 take reasonable steps to obtain on-site property damage insurance available at reasonable costs and at reasonable terms
-from private sources. The appropriate insurance has'been acquired and will be m intained for TMI on a site-basis.
50.54(x) and 50.54(y)
Paragraph 50.54(x) allows a licensee to take action which departs from a license condition or technical specification in an emergency when this action is
.immediately.needed to protect the health and safety of the public.. Paragraph 50.54(y) requires that any action taken pursuant to paragraph 50.54(x) be approved, as a minimum, by a senior operator prior to taking.the action.
The provisions of:
this. article have'. limited applicability to TMI-2 during PDMS..Due to the non-operating and defueled status of-TMI-2 during PDMS, there are no postulated events which could affect public health and safety _in such a manner.
In addition, the. technical specifications will be of limited scope and_ it is not anticipated that a condition will exist such that it.could become necessary to take action that 3.1-15 AMENDMENT 8 - OCTOBER 1990
l-f,}
l' departs 190mieither a license con'dition or a technical specification to protect public health and. safety.
Since THI-2 will not have licensed reactor operators L
during-PnMS, if,an extremely unlikely event were to occur necessitating deviation
'j
' from the-technical specifications the action would have to be approved by senior.
management.
50.54(z)-
Paragraph 50.54(z) requires each licensee to notify the plRC Operations Center of-i the occurrence of any event specified in 10 CFR 50.72.
Due to the non-operating 1
- and defueled status of THI-2 during PDMS, there are very few potential events which 1
would require. reporting under 10 CFR 50.72. However, to-the extent that reporting 1
is required under 10 CFR 50.72, the requirements of this paragraph are applicable.
i See Section 3.1.1.56 regarding paragraph 10 CFR 50.72.
l 50.54(aa) l Paragraph 50.54(aa) establishes that the licensee must meet Sections 401(a)(2) and
/
~401(d) of the Federal Water Pollution Control Act.
No exceptions are taken to the f
provisions of this article.
.j 50.54(bb)
Paragraph 50.54(bb) requires licensees of operating nuclear power reactors to f
acquire NRC approval of the program to fund, manage, and transfer irradiated fuel upon expiration of the reactor operating license.
The irrad W J fuel which j
comprised the TMI-2 reactor core has been transferred:to the possession of the:
j Department of Energy. The requirements of this paragraph have been satisfied.
50.54(cc)
Paragraph 50.54(cc) requires licensee written notifications of the appropriate NRC Regional Administrator of certain bankruptcy filings.
No exceptions are taken to-3 the provisions of this article.
1 1
50.~54(dd) i 1
Paragraph-50.54(dd) allows licensees to take reasonable actions that depart from a t
license condition or a-Technical' Specification under certain conditions during~a-National Security Emergency.
No exceptions are taken.to the provisions of this E
article.
p q
3.1.1.44 10 CFR 50.55 - Conditions of. construction permits.
]
Article 50.55 establishes terms and conditions of construction permits.
No exceptions are taken to the provisions of this article.
1 e
3.1-16 AMENDMENT 8 - OCTOBER 1990
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L 3.1.1=. 45 10 CFR 50.55a - Codes and standards.
2 -Article-50.55a requires each operating license for a-nuclear power facility be
~
subject to the inservice inspection _ requirements of paragraph (g) to Article 50.55a
'and that'each construction permit be subject to the remaining paragraphs of the
- article.- As this: application is not for a construction permit, paragraph (g) of iArticle 50.55a is the only' portion of the article potentially applicable to TMI-2 in PDMS.
Paragraph (g)'(1) is~the paragraph which applies to THI-2.
This paragraph requires that safety related components (including supports) meet the requirements of paragraph (g) (4) and (5).
These paragraphs define the in service inspection requirements and provides a mechanism for relief from impractical requirements based on a satisfactory demonstration of this action to the Nuclear Regulatory Commission. Relief from these requirements was sought and for the most part granted early in the TMI-2 Cleanup Program. The NRC granted the following-relief forTMI-2(Reference 3.1-5):
1.
The provisions of IWA-2400 of Section XI of the ASME Boiler and Pressure vessel = Code (code) 1974 edition, Summer 1975 Addenda for extending the inspection interval for a period of time equivalent to the shutdown period of THI-2 was applicable.
2.
_ Testing of pumps in accordance with Section IWP-3400 of the code was only required for those pumps specified in the Recovery Technical Specifications.
3.
Category A valves were defined to be containment isolation valves.
The NRC agreed that.these valves should not be exercised. However, 10 CFR Part 50 Appendix J Type "C" testing was required for any containment isolation valve that was opened and subsequently closed in order to verify its containment isolation function.
4.
Category B and C valves in systems out-of-service need not be tested, however, Category B and C valves;in safety related systems in service should be-exercised at least once per 92 days.
5.
The Mini Decay Heat Removal System was to be handled as a separate action.
During PDMS..the relief granted by the NRC still applies.
In addition, based on the following justification, no:further_ inservice inspection is required.
The PDMS Technical Specifications require no pumps to be operable. Therefore, 1..
based on the existing relief, testing of pumps in accordance with IWP-3400 would not-be required during PDMS.
2.
Performance of Type "C" testing of containment isolation valves (Category A-
, valves)-_in accordance with 10 CFR Part 50 Appendix J is not required at TMI-2.
The NRC granted THI-2-an u emption from Type "C" testing (Reference 3.1-4).
3.1-17 AMENDMENT 8 - OCTOBER 1990
n
}
4 r
j 3 'J During~PDMS,.therewillbenosafetyrelatcdsystemsatITMI-2. Therefore, j
performance of_ testing of Category 8 and C valves is not required based on
+
s
'the relief granted by the NRC.
i 24L No; testing is. required for the MDHR System as.it will be deactivated for
.PDMS:(seeSection6.25).
Therefore,ias discussed above, complete relief from the inservice _ inspection requirements of.10 CFR 50.55a during PDMS.is appropriate.
- 3.1.1.46 10 CFR 50.56 - Conversion of construction permit to license; or i
amt.ndment of license.
l Article ;0 36' establishes that the Commission will, in the absence of good _cause-1 shown to che contrary, issue a license or amendment of a license as the case may q
De.
No exceptians are taken to the provisions of this article.
3.1.1.47 10 rFR 50.57 - Issuance of operating license.
3 i
Article 50.57 establishes the standards the Commission shall use in determining the issuance of an operating license.
No exceptions are taken to the provisions 1
of this article.
4 13.1.1.48 10 CFR 50.58 - Hearings and report of the Advisory Committee on Reactor Safeguards.
4 50.58(a)
Paragraph 50.58(a) establishes that each application for-a construction permit, an operating-license, or an amendment to the construction permit or operating license J
may be referred to the Advisory Committee on Reactor Safeguards.
The report from.
the Advisory Committee on Reactor Safeguards will be made part of the public record.
No exceptions are taken to the provisions of this article.
50.58(b) 3 LParagraph 50.58(b) establishes-that the Commission may hold hearings on each
-application for a construction permit or an operating' license for a production or-utilization facility.of the type described in 10 CFR 50.21(b) or 10 CFR 50.22. No-exceot Mns are taken to the provisions of this paragraph.
. 3.1'.1. 4 9 10 CFR 50.59 - Changes, tests and experiments.
Article 50.59 establishes the requirements for changes, tests or experiments which
-affect 1the' facility. ~ TMI-2 will address this article by evaluating facility
- changes, tests, or experiments utilizing the intent of the guidelines given in 10 "CFR 50.59.. For example, any activity which could make sufficient fuel mobile to.
finvalidate:or bring.into question the validity of the criticality analysis as' j
presented in SAR.Section 4.3 would require NRC review under the guidelines of 10 1
- CFR'50.59.
3.1-18 AMEN 0 MENT 8 - OCTOBER 1990 e
t.
a,
Y N".1.1.501 10.CFR 50.60 - Acceptance criteria for fracture prevention measures for=
lightwater nuclear power reactors for normal operation.
10 CFR 50.60 requires all light water. nuclear _ power reactors to meet the fracture toughness and material surveillance program requirements for the reactor coolant-pressure boundary as_ set forth in Appendices G and H to 10 CFR 50.
10 CFR 501 Appendix G " Fracture Toughness Requirements" specifies fracture toughness requirements for ferretic materials of-pressure retaining components of the reactor
~
coolant pressure boundary of light water cooled reactors to provide adequate margins of safety during any condition of normal operation. With the completion of TMI-2 defueling and the draining of the Reactor Coolant System, the need for-meeting these requirements no longer exists. With the TMI-2 reactor defueled, normal operation is no longer possible and draining of the Reactor Coolant System precludes it from having to act as a pressure boundary.
Thus, performance of the testing program and compliance with the requirements specified in this appendix is no_ longer necessary.
Similarly,10 CFR 50 Appendix H " Reactor Vessel Material Surveillance Program Requirements" specifies material surveillance program criteria for monitoring-fracture toughness properties of ferretic material in the Reactor Vessel beltline region. The purpose of the program is to determine how these properties change due to neutron irradiation and the thermal environment, and to demonstrate that the-fracture toughness requirements specified in 10 CFR 50-Appendix G continue to be met. -With the completion of defueling of the-THI-2 reactor, neutron embrittlement-of the Reactor Vessel is no longer a concern at THI-2, and, consequently there is-no longer a need to continue this program.
Application of this regulation would not serve the underlying purpose of the-rule
-which is to provide an adequate margin of safety from failure of the reactor coolant pressure boundary during any condition of normal operation, including anticipated operational. occurrences and. system-hydrostatic tests. Thus, in accordance with 10 CFR 50.12(a)(2)(ii) exempting TMI-2 from this regulation during PDMS is appropriate.
3.1.1.51 10 CFR 50.61 - Fracture toughness requirements for protection _against pressurized thermal shock events.
Article 50.61 establishes ~ requirements for the protection against pressurized thermal shock in pressurized water nuclear power reactors. Due to the Lnon-operating and defueled status of THI-2 during PDMS, the requirements of this article do not apply.. In addition, TMI-2 was granted an exemption to 10 CFR 50.61
_(Reference 3.1-1)whichacknowledgedthatTHI-2needtakenomeasurestoprotect against pressurized thermal shock.
3.1.1.52 10 CFR 50.62 - Requirements for reduction of risk from anticipated transients without scram (ATWS) events for light-water-cooled nuclear power reactors.
3.1-19 AMENDMENT 8 - OCTOBER 1990 l
W.i Article 50.62 establishes requirements for comercial light-water-cooled nuclear 1 Lpower plants to.have equipment.to address ATWS events. An ATWS event is a.
p'
. condition of normal-operation. expected to'. occur one or more times during the life
-of a facility:followed by a failure of the reactor trip portion of the protection 4
system.-- Due to the non-operating and defueled status of TMI-2 during PDMS, TMI-2 Nis incapable of normal operation.
Thus, the requirements of this' article do not-
- apply.:
3.1.1.53 10 CFR 50.63 - Loss-of alternating current power.
" I-
. Article 50.63 requires that each light-water-cooled nuclear power plant licensed to operate be able to withstand'and recover from a station blackout event.
Since this
- application eliminates the legal authority to operate the TMI-2 facility from the license, a subsequent license application would be necessary to resume. operation.-
Therefore, during PDMS the requirements of this article are not applicable.to TMI-2.
c.
3.1.1.54 10 CFR 50.64 - Limitations on the use of highly enriched uranium-(HEU) l in domestic non-power reactors.
-Article 50.64 establishes requirements for the issuance of licenses to use highly enriched unranium fuel in non-power reactors.
No exceptions are taken to the provisions'of this article.
-3.1.1.55 10 CFR-50.70 - Inspections.
Article 50.70 establishes requirements to permit NRC inspectors to maintain activities at each nuclear power plant site. During PDMS, TMI-2 will be required to support NRC inspection activities to the extent determined necessary by the NRC.
No exceptions are taken to the provisions of this article.
3.1.1'.56 10 CFR 50.71'- Maintenance of records, making of reports.
y
"~
50.71(a) through 50.71(d)
Paragraphs 50.71(a) through 50.71(d) establish requirements.for facility records.
!The requirements of these paragraphs. apply to TMI-2 during PDMS.
.V 50.71(e) t Paragraph 50.71(e) establishes that each person licensed to operate a nuclear power reactor.shall periodically update the FSAR. TMI-2 will-not be licenseo to operate
'a nuclear power reactor; therefore, the requirements 'of-this paragraph do not applyi
~
to TMI-2 during PDMS. However, the intent of this paragraph has been addressed by y
the inclusion of any information regarding revisions to the PDMS SAR in the annual report specified.in Technical Specification 6.7.1.2.
3.1-20 AMENDMENT 8 - OCTOBER 1990 s
1
'3.1.1.57 10 CFR 50.72 - Immediate notification requirements for operating l
nuclear ponte reactors.
With the exception of paragraphs 50.72(b)(2)(iv)(A), 50.72(b)(2)(iv)(B),
j 50.72(b)(2)(v), and 50.72(b)(2)(vi) the requirements for notification address events or s: n' ions which are related to the operation of the power plant and conditions w,
.a do, or may compromise the safe operation of the plant.
Since I
THI-2 will be specifically precluded from the operation of the plant during PDMS, the requirements of those paragraphs which relate to power plant operation will not apply.
\\
Paragraphs 50.72(b)(2)(iv)(A) and 50.72(b)(2)(iv)(B) establish requirements for reporting liquid and ghseous effluents which exceed levels established by these 1
paragraphs.
These requirements are applicable to TMI-2 during PDMS.
Paragraphs 50.72(b)(2)(v) and 50.72(b)(2)(vi) require the reporting of any event requiring the transport of a radioactively contaminated pers m to an off-site i
medical facility for treatment and any event or situation, sted to the health and safety of the public or onsite personnel, or protection the environment, for which a news releast is planned or notification to other vernment agencies has been or will be made.
These requirements are also appli' 21e to TMI-2 during PDMS.
3.1.1.58 10 CFR 50.73 - Licensee event report system.
Article 50.73 requires that the holder of an operating license for nuclear power plant (licensee) shall submit a Licensee Event Report (LER) for any event of the type described in this paragraph within 30 days after the discovery of the event.
Since TMI-2 during PDMS is specifically precluded from the operation of the plant, the requirements of this paragraph do not apply.
In addition, the " statements of consideration" for this rule (48 FR 33850) states the following:
"This rule identifies the types of reactor events and problems that are believed to be significant and useful to the NRC in its effort to identify and resolve threats to public safety.
It is designed to provide the information necessary for engineering studies of operational anomalics and trends and patterns analysis of operational occurrences. The same information can also be used for other analytic procedures that will aid in identifying accident precursors."
The statement of consideration for this rule clearly specifies that the purpose of this rule is to accumulate information from operational occurrences. Due to the non-operating and defueled status of TMI-2 during PDMS, the facility clearly does not fall within the stated objectives of this rule.
3.1-21 AMENDMENT 8 - OCTOBER 1990
I.1.1.59 10 CFR 50.75 - Reporting and recordkrping for d:comissicning planning.
Article 50.75 establishes requirements for providing reasonable assurance to the NRC that funds will be available for decomissioning. No exceptions are taken to
-the provisions of this article. Additionally, Reference 3.1-6 provided the decomissioning funding plan for THI-2.
3.1.1.60 10 cfR 50.78 -Installation information and verification.
l Arti:le 50.78 requires that, "Each holder of a construction permit shall, if requested by the Comission, submit ins'allation information on Form N-71, permit verification thereof by the International Atomic Energy Agency, and take such other action as may be necessary to implement the US/lAEA Safeguards Agreement, in the manner set forth in Articles 75.6 and 75.11 through 75.14 of this chapter."
No exceptions are taken to the provisions of this article.
~
3.1.1.61 10 CFR 50.80 - Transfer of licenses, j
Article 50.80 specifies requirements for transferring a license from one ent!'y to another.
No exceptions are taken to the provisions of this article, j
3.1.1.62 10 CFR 50.81 - Creditor regulations.
l Article 50.81 defines the rights and restrictions applying to any creditor relative to any license issued by the Commission.
No exceptions are taken to the provisions of this article.
3.1.1.63 10 CFR 50.82 - Applications for termination of licenses.
l Article 50.82 defines the requirements for terminating a license.
No exceptions are taken to the provisions of this article.
3.1.1.64 10 CFR 50.90 - Application for amendment of license or construction l
permit.
Article 50.90 establishes that a holder of a license must file an application for an amendment, describing the changes desired, if the license holder wishes to amend the license.
No exceptions are taken to the provisions of this article.
3.1.1.65 10 CFR 50.91 - Notice for public comment; state consultation.
Article 50.91 establishes requirements applying to the Comission and THI-2 regarding the application for an amendment to an operating license.
This article requires that the licensee must file a no significant hazards consideration analysis using the standards set forth in Article 50-92 when filing a license amendment application. The article then establishes the Comissions allowed actions depending 'on the determination of no significant hazards.
The requirements of this article apply to TMI-2.
i L
3.1-22 AMENDMENT 8 - OCTOBER 1990
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7
'3.1.1.66 10 CFR 50.92 - Issuance of amenament.
l Article 50.92 establishes the standards by which the Comission determines if no significant hazards exist for a license amendment.
The licensee must file a no significant hazards analysis with each amendment application using the standards set forth in Article 50.92 as required by Article 50.91.
The requirements of this article apply to TM1-2.
3.1.1.67 10 CFR 50.100 - Revocation, suspension, modification, of licenses and l
construction permits for cause.
Article 50.100 provides that the Comission may revoke, suspend, or modify a license or construction permit for any material f alse statement or for other reasons specified in Article 50.100.
No exceptions are taken to the provisions of this article.
3.1.1.68 10 CFR 50.101 - Retaking possession of special nuclear material.
l i
Article 50.101 establishes that the Comission may cause the retaking or possession of special nuclear material upon revocation of a license. No exceptions are taken to the provisions of this article.
3.1.1.69 10 CFR 50.102 - Comission order for operation after revocation.
l Article 50.102 establishes that the Comission may, by following the requirements of Article 50.102, order operation of a facility whose license has been revoked.
No exceptions are taken to the provisions of this article.
i 3.1.1.70 10 CF3 50.103 - St.spension and operation in war or national emergency.
l Article 50.103 establishes that the Comission has, upon declaration of war by the l
l Congress, certain rights regarding the suspension and/or operation of nuclear power plants tiicensed by the Comission. No exceptions are taken to the provisions of this article.
l 3.1.1.71 10 CFR 50.109 - Backfitting.
l Article 50.109 defines backfitting and defines requirements the Comission must meet regarding backfitting. No exceptions are taken to the provisions.of this article.
3.1.1.72 10 CFR 50.110 - Violations, g
Article 50.110 establishes actions the NRC may take regarding violations of any i
provision of the Atomic Energy Act of 1954, as amended, or Title II of the Energy j
Reorganization Act of 1974, or any regulation or order issued thereunder.
No j
exceptions are taken to the provisions of this article.
l 1
3.1-23 AMENDMENT 8 - OCTOBER 1990
l
.3.1.2 GENERAL DESIGN CRITERIA The Three Mile Island Nuclear Station Unit 2 was designed and constructed in accordance with the 70 general design criteria as listed in Appendix A of 10 CFR 50 dated July 11, 1967. A discussion of each criterion, demonstrating how the 4
principal design features or design bases meet these criteria, is presented in Section 3.1.1 of the TH1-2 FSAR.
The general design criteria in Appendix A were revised by the AEC on July 15, 1971.
The design and purchase of many Three Mile Island Unit 2 components were completed prior to the issuance of these revised general design criteria.
These revised criteria, as they applied to the original design of the plant, are addressed in Section 3.1.2 of the THI-2 FSAR.
During the PDMS period, fulfillment of many of the general design criteria in Appendix A of 10 CFR 50 are not necessary or appropriate; departure from the criteria are identified and justified herein. Other of the criteria are applicable only to a very limited degree. Criteria which address such requirements as containment, quality standards, and natural phenomena are examples of those criteria which apply only to a limited degree during PDMS.
Since the plant was originally designed and constructed in accordance with these criteria and since neither the occident nor activities during the recovery period significantly degraded the plant with respect to the capabilities required during PDMS, the facility, as it exists, is designed and constructed to standards which far exceed the requirements for PDMS.
Each of the general design criteria in Appendix A of 10 CFR 50, as revised on January 1,1987, and the necessary and appropriate degree of applicability during PDMS is discussed in the following sections.
3.1.2.1 Criterion 1 - Quality standards and records.
Structures, systems, and components important to safety shall be designed, I
fabricated, erected, and tested to quality standards commensurate with the importance of the safety function to be performed. Where generally recognized codes and standards are used, they shall be identified and evaluated to determine their applicability, adequacy, and sufficiency and shall be supplemented or modified as necessary to assure a quality product in keeping with the required safety function.
A quality assurance program shall be established and 5plemented in order to provide adequate assurance that these structures, systems, and components will satisfactorily perform their safety functions. Appropriate records of the design, fabrication, erection, and testing of structures, systems, and components important to safety shall be maintained by or under the control of.the nuclear power unit licensee throughout the life of the unit.
3.1-24 AMENDMENT 8 - OCTOBER 1990
'Discussirn Due to the unique condition of TMI-2 during PDMS, the specific requirements of Criterion 1 are not applicable; however, the intent of Criterion I has been addressed recognizing that the degree of quality assurance necessary to assure that the required capabilities are maintained during PDMS is far less extensive than that which was originally required for THI-2. A quality assurance program has been established and will be maintained cummenturate with the functional requirements of PDMS.
The Quality Assurance Plan for PMS 4 refc enced in Section 10.1.
3.1.2.2 Criterion 2 - Desigr bases for protection against natural phenomena.
Structures, systems, and components important to safety shall be designed to withstand the effect of natural phenomena, such as earthquakes, tornadoes,
(
hurricanes, floods, tsunami, and seiches without loss of capability to perform their safety functions. The design bases for these structures, systems, and components shall reflect:
(1) Appropriate consideration of the most severe of the natural phenomena that have been histo-ically reported for the site and surrounding area, with sufficient margin for the 1 mited accuracy, quantity, and period of time in which the historical data have been accumulated, (2) appropriate combinations of the efftets of normal and accident conditions with the effects of the natural phenomena and (3) the importance of the safety functions to be performed.
Discussion Due to the unique condition of THI-2 during PDMS, the specific requirements of Criterion 2 are not applicable; however, the intent of Criterion 2 has been addressed by recognizing that the level of protection from natural phenomena required during PDMS is that which is required t maintain the isolation of the contamination which remains at the facility. TL re are no active functions l
required to be performed by any system to provide the protection from natural phenomena during PDMS.
For example, all that is required during a seismic event is that the structure or system remain intact. Those structures, systems, and components necessary for the level of protection required for PDMS were originally designed and constructed to criteria which exceed the requirements for PDMS. This j
level of protection is more than adequate to meet the functional requirements for l
protection from natural phenomena during PDMS.
3.1.2.3 Criterion 3 - Fire protection.
Structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and
)
effect of fires and explosions. Noncombustible and heat resistant materials shall l
be used whenever practical throughout the unit, particularly.in locations such as the containment and control room.
Fire detection and fighting systems of appropriate capacity and capability shall be provided and designed to minimize the adverse effects of fires on structures, systems, and components important to safety.
Firefighting systems shall be designed to assure that their rupture or inadvertent operation does not significantly impair the safety capability of these structures, systems, and components.
3.1-25 AMENDMENT 8 - OCTOBER 1990
D'iscussien
~'
Due to the unique condition of THI-2 during PDMS, the specific requirements of Criterion 3 are not applicable; however, the intent of Critarion 3 has been addressed by recognizing that the requirements for fire protection during PDMS are based on industrial safety and insurance requirements. A fire protection pr(gram has been established and will be maintained commensurate with the industrial safety and insurance requirements and to protect those systems important to PDMS. The Fire Protection System is described in Section 7.2.2.
3.1.2.4 Criterion 4 - Environmental and missile design bases.
Structures, systems, and components important to safety shall be designed to accommodate the effects of and be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents. These structures, systems, and components shall be appropriately protected against dynamic effect, including the effects of missiles, pipe whipping, and discharging fluids, that may result from equipment failures and from events and conditions outside the nuclear power unit.
- However, the dynamic effects associated with postulated pipe ruptures of primary coolant loop piping in pressurized water reactors may be excluded from the design basis when analyses demonsteate the probability of rupturing such piping is extremely low under design basis conditions.
Discussion Due to the non-operating and defueled condition of TMI-2, the requirements of Criterion 4 associated with the dynamic effect, including the effects of missiles, pipe whipping, and discharging fluids, that may result from equipment failures and from events and conditions outside the nuclear power unit do not apply.
Structures, systems and components relied upon to provide protection from the effects of and required to be compatible with environmental conditions associated with PDMS operations, maintenance, testing, and postulated unanticipated events are appropriately designed to accommodate effects associated with these activities.
3.1.2.5 Criterion 5 - Sharing of structures, systems, and components.
Structures, systems, and components important to safety shall not be shared among nuclear power units unless it can be shown that such sharing will not significantly impair their ability to perform their safety function, including, in the event of an accident in one unit, an orderly shutdown and cooldown of the remaining units.
Discussion Due to the non-operating and defueled condition of TMI-2, there are no important to safety functions associated with any THI-2 structure, system, or component. The
. required TMI-1 safety functions associated With the few structures and systems shared by THI-1 and THI-2 are independent of any TMI-2 function for the respective structure or system.
3.1-26 AMENDMENT 8 - OCTOBER 1990
l l
3'.1. 2. 6 Critericn 10 - Reactor design.
The reactor core and associated coolant, control, and protection systems shall be
. designed with appropriate margin.to assure that specified acceptable fuel design limits are not exceeded during any condition of nomal operation, including the effects of anticipated operational occurrences.
Discussion s.a to the non-operating and defueled condition of TMI-2, the requirements of Criterion 10 are not applicable.
3.1.2.7 Criterion 11 - Reactor inherent protection.
l The reactor core and associated coolant systems shall be designed so that in the power operating range the net effect to the prompt inherent nuclear feedback characteristics tends to compensate for a rapid increase in reactivity.
Discussion Due to the non-operating and defueled condition of THI-2, the requirements of Criterion 11 are not applicable.
3.1.2.8 Criterion 12 - Suppression of reactor power oscillations.
The reactor core and associated coolant, control, and protection systems shall be
' designed to assure that power oscillations which can result in conditions exceeding
.specified acceptable fuel design limits are not possible or can be reliably and-readily detected and suppressed.
Discussion Due to the non-operating and defueled condition of THI-2, the requirements of Criterion 12 are-not applicable.
3.1.2.9 Criterion 13 - Instrumentation and control.
Instrumentation shall' be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety, including those variables and systems that can affect the fission process, the integrity of the; reactor core, the reactor coolant pressure boundary, and the containment and its associated systems.
Appropriate controls shall be provided to maintain these variables and systems within prescribed operation ranges.
Discussion-Due to the non-operating and defueled condition of THI-2, the requirements of Criterion 13 are not applicable.
3.1-27 AMENDMENT 8 - OCTOBER 1990
l 3.1.2.10 Critoritn 14 - React:r coolant pressure boundary.
The reactor coolant pressure boundary shall be designed, fabricated, erected, and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture.
Discussion Due to the non-operating and defueled condition of THI-2, the requirements of Criterion 14 are not applicable.
3.1.2.11 Criterion 15 - Reactor coolant system design.
The reactor coolant system and associated auxiliary, control, and protection systems shall be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during any condition of normal operation, including anticipated operational occurrences.
Discussion Due to the non-operating and defueled condition of THI-2, the requirements of Criterion 15 are not applicable.
3.1.2.12 Criterion 16 - Containment design.
Reactor containment and associated systems shall be provided to establish an
-essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment and to assure that the containment design conditions important to safety are not exceeded for as long as postulated accident conditions require.
Discussion l
The Containment and associated systims are maintained during oDMS to prevent t.;
uncontrolled release of the contamiration which remains inside the Containment.
In addition, the Containment serves as che primary environmental shielding of the radioactive materials inside the Contiinment.
Although the Containment will not be maintained during PDMS to the same degree of leaktightness as during power operation, there will be essentially no uncontrolled leakage. Normally, all effluents to the environment will be.through the the Containment Atmospheric l
Breather System via the Auxiliary Building or the Containment Purge System, both of.
which have HEPA filter systems.
Leakage to the environment from other pathways has i
been demonstrated, by analysis (see Section 7.2.1.2), to be a very small portion of l
the overall.-leakage from the Containment. The Containment Atmospheric Breather System controls the Containment' effluents during passive storage periods and is consistent with the "most probable pathway"_ concept referred to in Regulatory Guide 1.86 " Termination of Operating. License for Nuclear Reactors."
3.1-28 AMENDMENT 8 --0CTOBER 1990 i
3.1.2.13 Criterien 17 - Electric power systems.
An onsite electric power system and an off-site electric power system shall be provided to permit functioning of structures, systems, and components important to safety. The safety function for each system (assuming the other system is not functioning) shall be to provide sufficient capacity and capability to assure that (1) specified acceptable fuel design limits and design conditions of the reactor i
coolant pressure boundary are not exceeded as a result of anticipated operational occurrences and (2) the core is cooled and containment integrity and other vital functions are maintained in the event of postulated accidents.
The onsite electric power supplies, including the batteries, and the onsite electric distribution system, shall have sufficient independence, redundancy, and testability to perform their safety functions assuming a single failure.
Electric power from the transmission network to the onsite electric distribution system shall be supplied by two physically independent circuits (not necessarily on separate rights of way) designed and located so as to minimize to the extent practical the likelihood of their simultaneous failure under operating and postulated accident and environmental conditions.
A switchyard common to both circuits is acceptable.
Each of these circuits shall be designed to be available in sufficient time following a loss of all onsite alternating current power supplies and the other off-site electric power circuit, to assure that specified acceptabic fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded. One of these circuits shall be designed to be available within a few seconds following a loss-of-coolant accident to assure that core cooling, containment integrity, and other vital safety functions are maintained.
Provisions shall be included to minimize the probability of losing electric power from any of the remaining supplies as a result of, or coincident with, the loss of power generated by the nuclear power unit, the loss of power from the transmission network, or the loss of power from the onsite electric power supplies.
Discussion Due to the non-operating and defueled condition of TMI-2, the requirements of Criterion 17 are not applicable. However, capabilities for electric power are maintained during PDMS commensurate with the electric power requirements necessary for PDMS activities.
3.1.2.14 Criterion 18 - Inspection and testing of electric power systems.
Electric power systems important to safety shall be designed to permit appropriate periodic inspection and testing of important areas and features, such as wiring, insulation, connections, and switchboards, to assess the continuity of the systems and the condition of their components. The systems shall be designed with a capability to test periodically (1) the operability and functional performance of the components _of the systems, such as onsite power sources, relays, switches, and buses, and (2) the operability o,f the systems as a whole and, under conditions as close to design as practical, the full operation sequence that brings the systems 3.1-29 AMENDMENT 8 - OCTOBER 1990
\\'
1nto cperation, including operation of applicable portions of the prctecticn system, and the transfer of power among the nuclear power unit, the off-site power system, and the onsite power system.
Discussion Due to the non-operating and defueled condition of THI-2, the requirements of Criterion 18 are not applicable.
3.1.2.15 Criterion 19 - Control room.
A control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including loss-of-coolant accidents.
Adequate radiation protection.shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident.
Equipment at appropriate locations outside the control room shall be provided (1) with a design capability for prompt hot shutdown of the reactor, including necessary instrumentation and controls to maintain the unit in a cafe condition during hot shutdown, and (2) with a potential capability for subsequent cold shutdown of the reactor through the use of suitable procedures.
Discussion i
Due to the non-operating and defueled condition of THI-2, the requirements of Criterion 19 are not applicable.
3.1.2.16 Criterion 20 - Protection system functions.
The protection system shall be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operation occurrences and (2) to sense accident conditions and to initiate the operation of systems and components important to safety.
Discussion Due to the non-operating and defueled condition of THI-2, the requirements of Criterion 20 are not applicable.
3.1.2.17 Criterion 21 - Protection system reliability and testability.
The prote:: tion system shall be designed for high functional reliability and inservice testability commensurate with the safety functions to be performed.
Redundancy and independence designed into the protection system shall be sufficient to assure that (1) no single failure results in loss of the protection a
function and (2) removal from service of any component or channel does not result l
in: loss.of the required minimum redundancy unless the acceptable reliability of 3.1-30 AMENDMENT 8 - OCTOBER 1990
i 1
bperati n of the protection system can be otherwise demonstrated. The protecticn system shall be designed to permit periodic testing of its functioning when the reactor is in operation, including a capability to test channels independently to determine failures and losses of redundancy that may have occurred.
1 Discussion Due to the non-operating and defueled condition of THI-2, the requirements of Criterion 21 are not applicable.
3.1.2.18 Criterion 22 - Protection system independence.
4 The protection system shall be designed to assure that the effects of natural phenomena, and of normal operating, maintenance, testing, and postulated accident conditions on redundant channels do not result in loss of the protection function, or shall be demonstrated to be acceptable on some other defined basis.
Design techniques, such as functional diversity or diversity in component design and prim:1ples of operation, shall be used to the extent practical to prever.t loss of the protection function.
Discussion Due to the non-operating and defueled condition of TMI-2, the requirements of Criterion 22 are not applicable.
3.1.2.19 Criterion 23 - Protection system failure modes.
The protection system shall be designed to fail into a safe state or into a state demonstrated to be acceptable on some other defined basis if conditions such as disconnection of the system, loss of energy (e.g., electric power, instrument air), or postulated adverse environments (e.g., extreme heat or cold, fire, pressure, steam, water, and radiation) are experienced.
t Discussion Due_to the non-operating and defueled condition of THI-2, the requirements of-Criterion 23 are not applicable.
3.1.2.20 Criterion 24 - Separation of protection and control systems.
The protection system shall be separated from control systems to the extent that failure of any single control system component or channel, or failure or removal from service of any single protection system component or channel which is common to the control and protection systems leaves intact a system satisfying all reliability, redundancy, and independence requirements of the protection system.
Interconnection of the protection and control systems shall be limited so as to assure that safety is not significantly impaired.
3.1-31 AMENDMENT 8 - OCTOBER 1990 L
l Iiscussirn Due to the non-operating and defueled condition of TMI-2, the requirements of Criterion 24 are not applicable.
3.1.2.21 Criterion 25 - Protection system requirements for reactivity control malfunctions.
The protection system shall be designed to assure that specified acceptable fuel design limits are not exceeded for any single malfunction of the reactivity control systems, such as accidental withdrawal (not ejection or dropout) of control rods.
Discussion Due to the non-operating and defueled condition of THI-2, the requirements of Criterion 25 are not applicable.
3.1.2.22 Criterion 26 - Reactivity control system redundancy and capability.
Two independent reactivity control systems of different design principles shall be provided. One of the systems shall use control rods, preferably including a positive means for inserting the rods, and shall be capable of reliably controlling reactivity changes to assure that under conditions of normal operation, including anticipated operational occurrences, and with appropriate margin for malfunctions such as stuck rods, specified acceptable fuel design limits are not exceeded.
The second reactivity control system shall be capable of reliably controlling the rate of reactivity changes resulting from planned, normal power changes (including xenon burnout) to assure acceptable fuel design limits are not exceeded. One of the systems shall be capable of holding the reactor core subcritical under cold conditions.
Discussion Due to the non-operating and defueled condition of THI-2, the requirements of
-Criterion 26 are not applicable.
3.1.2.23 Criterion 27 - Combined reactivity control systems capability.
The reactivity control syst m shall be aesigned to have a combined capability, in conjunction with poison ar.; tion by the emergency core cooling system, of reliably controlling reactivity changes to assure that under postulated accident conditions
-and with appropriate margin for stuck rods the capability to cool the core is j
maintained.
Discussion Due to the non-operating and defueled condition of TMI-2, the requirements of Criterion 27 are not applicable.
3.1-32 AMENDMENT 8 - OCTOBER 1990 L
t I.1.2.24 Criterion 28 - Reactivity liaits.
The reactivity control systems shall be designed with ar.opriate limits on the potential amount and rate of reactivity increase to assure that the effects of postulated reactivity accidents can neither (1) result in damage to the reactor
. coolant pressure boundary greater than limited local yielding nor (2) sufficiently disturb the core, its support structures of other reactor pressure vessel internals to impair significantly the capability to cool the core. These l
postriated reactivity accidents shall include consideration of rod ejection (unless prevented by positive means), rod dropout, steam line rupture, changes in recctor coolant temperature and pressure, and cold koter addition.
Discussion Due to the non-operating and defueled condition of THI-2, the requirements of Criterion 28 are not applicable.
3.1.2.25 Criterion 29 - Protection against anticipated operational occurrences.
The protection and reactivity control systems shall be designed to assure an extremely high probability of accomplishing their safety functions in the event of anticipated operational occurrences.
Discussion Due to the non-operating and defueled condition of TMI-2, the requirements of Criterion 29 are not applicable.
3.1.2.26 Criterion 30 - Quality of reactor coolant pressure boundary.
Components which are part of the reactor coolant pressure boundary shall be designed, fabricated, erected, and tested to the highest quality standards practical. Means shall be provided for detecting and, to the extent practical, identifying the location of the source of reactor coolant leakage.
Discussion Due to the non-operating and defueled condition of THI-2, the requirements of Criterion 30 are not applicable.
l 3.1.2.27 Criterion 31 - Fracture prevention of reactor coolant pressure boundary.
L The reactor coolant pressure boundary shall be designed with sufficient margin to assure that when stressed under operating. maintenance, testing, and postulated accident conditions (1) the boundary behaves in a nonbrittle manner and (2) the probability of rapidly propagating fracture is minimized. The design shall reflect consideration of service temperatures and other conditions of the boundary l
material under operating, maintenance, testing, and postulated accident conditions and the uncertainties in determining (1) material properties, (2) the effects of irradiation of material properties, (3) residual, steady state and transient' stresses, and (4) size of flaws.
3.1-33 AMENDMENT 8 - OCTOBER 1990
t)'iscussion o.
Due to the non-operating and defueled condition of TMI-2, the requirements of Criterion 31 are not applicable.
-3.1.2.28 Criterion 32 - Inspection of reactor coolant pressure boundary.
Components which are part of the reactor coolant pressure boundary shall be l
designed to permit (1) periodic inspection and testing of important areas and i
features to assess their structural and leaktight integrity, and (2) an appropriate material surveillance program for the reactor pres m re vessel.
Discussion Due to the non-operating and defueled condition of THI-2, the requirements of Criterion 32 are not applicable.
3.1.2.29 Criterion 33 - Reactor coolant makeup.
I A system to supply reactor coolant makeup for protection against small breaks in the reactor coolant pressure boundary shall be provided.
The system safety function shall be to assure that specified acceptable fuel design limits are not exceeded as a result of reactor coolant loss due to leakage from the reactor coolant pressure boundary and rupture of small piping or other small components which are part of the boundary.
The system shall be designed to assure that for onsite electric power system operation (assuming off-site power is not available) and for off-site electric power system operation (assuming onsite power is not available) the system safety function can be accomplished using the piping, pumps, and valve used to maintain coolant inventory during normal reactor operation.
1 1
Discussion Due to the non-operating and defueled condition of TMI-2, the requirements of Criterion 33 are not applicable.
3.1.2.30 Criterion 34 - Residual heat removal.
l A system to remove residual heat shall be provided. The system safety function l.
shall be to transfer fission product decay heat and other residual heat from the l
L reactor core at a rate such that specified acceptable fuel design limits and the design conditions of the reactor coolant pressure boundary are not exceeded.
l Suitable redundancy in components and features, and suitable interconnections, l
1eak detection, and isolation capabilities shall be provided to assure that for l
onsite electric power system operation (assuming off-site power is not available) and for off-site electric power system operation (assuming onsite power is not available) the system safety function can be accomplished,' assuming a single failure.
L 3.1-34 AMENDMENT 8 - OCTOBER 1990
l tiiscussion Due to the non-operating and defueled condition of TMI-2, the requirements of 1
Criterion 34 are not applicable.
3.1.2.31 Criterion 35 - Emergency core cooling.
A tystem to provide abundant emergency core cooling shall be provided.
The system safety function shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented and (2) clad metal-water reaction is limited to negligible amounts.
Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities shall be provided to assure that for onsite electric power system operation (assuming off-site power is not available) and for off-site electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.
Discussion Due to the non-operating and defueled condition of THI-2, the requirements of Criterion 35 are not applicable.
3.1.2.32 Criterion 36 - Inspection of emergency core cooling system.
The emergency core cooling system shall be designed to permit appropriate periodic inspection of important components, such as spray rings in the reactor pressure i
vessel, water injection nozzles, and piping, to assure the integrity and capability of the system.
Discussion s
Due to the non-operating and defueled condition of TMI-2, the requirement? of Criterion 36 are not applicable.
3.1.2.33 Criterion 37 - Testing of emergency core cooling system.
The emergency core cooling system shall be designed to permit appropriate periodic pressure and functional testing to assure (1) the structural and leaktight integrity of its components, (2) the operability and performance of the active i
components of the system, and (3) the operability of the system as a whole and, under conditions as close to design as practical, the performance of the full operational sequence that brings the system into operation, including operation _of l'
applicable portions of the protection system, the transfer between normal and emergency power sources, and the operation of the associated cooling water systt:n.
l 3.1-35 AMENDMENT 8 - OCTOBER 1990 l
t1 l
t)iscussirn
\\
Due to the non-operating and defueled condition of THI-2, the requirements of Criterion 37 are not applicable.
3.1.2.34 Criterion 38 - Containment heat removal.
l A system to remove heat from the reactor containment shall be provided.
The system safety function shall be to reduce rapidly, consistent with the functioning
)
of other associated systems, the containment pressure and temperature following any loss-of-coolant accident and maintain them at acceptably low levels.
Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities shall be provided to assure that for onsite electric power system operation (assuming off-site power is not available) and for off-site electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.
Discussion Due to the non-operating and defueled condition of TMI-2, the requirements of Criterion 38 are not applicable.
3.1.2.35 Criterion 39 - Inspection of containment heat removal system.
The containment heat removal system shall be designed to permit appropriate periodic inspection of important components, such as the torus, sumps, spray nozzles, and piping to assure the integrity and capability of the system.
Discussion Due to the non-operating and defueled condition of TMI-2, the requirements of Criterion 39 are not applicable.
~3.1.2.36 Criterion 40 - Testing of containment heat removal system.
The containment heat removal system shall be designed to permit appropriate periodic pressure and functional testing to assure (1) the structural and leaktight integrity of its components, (2) the operability and performance of the active components of the system, and (3) the operability of the system as a whole, l'
and under conditions as close to the design as practical the performance of the full operational sequence that brings the system into operation, including operation of applicable portions of the protection system, the transfer between i
normal and emergency power sources and the operation of the associated cooling L
. water system.
1; Discussion
~
Due to the non-operating and defueled condition of TMI-2, the requirements of Criterion 40 are-not applicable.
3.1-36 AMENDMENT 8 - OCTOBER 1990
.[.1.2.37 Criterion 41 - Containment atmosphere cleanup.
Systems to control fission product, hydrogen, oxygen, and other substances which may be released in the reactor containment shall be provided as necessary to reduce, consistent with the functioning of other associated systems, the h
concentration and quality of fission products released to the environment following postulated accidents, and to control the concentration of hydrogen or oxygen and other substances in the containment atmosphere following postulated accidents to assure that containment integrity is maintained.
Each system shall have suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities to assure that for onsite electric power system operation (assuming off-site power is not available) and for off-site electric power system operation (assuming onsite power j
is not available) its safety function can be accomplished, assuming a single failure.
U J
Discussion Due to the non-operating and defueled condition of THI-2, there are no postulated accidents during PDMS which could result in the generation of fission products, hydrogen, oxygen or other substances which would require Containment atmosphore cleanup systems as described in Criterion 41.
Therefore, design of the Containment atmosphere cleanup system for THI-2 during PDMS in accordance with Criterion 41 is not applicable.
See the analysis in Section 8.2.
3.1.2.30 Criterion 42 - Inspection of containment atmosphere cleanup systems.
The containment atmosphere cleanup systems shall be designed to permit appropriate periodic inspection of important components, such as filter frames, ducts, and piping to assure the integrity and capability of the systems.
Discussion Due to the non-operating and defueled condition of TMI-2, there are no postulated accidents during PDMS which could result in the generation of fission products, hydrogen, oxygen or other substances which would require containment atmosphere I
cleanup systems as described in Criterion 41. Therefore, design of the Containment atmosphere cleanup system in accordance with Criterion 42 is not applicable to TM1-2 during PDMS.
3.1.2.39 Criterion 43 - Testing of containment atmosphere cleanup systems.
1 i
The containment atmosphere cleanup systems shall be designed to permit appropriate
-periodic pressure and functional testing to assure (1) the structural and leaktight integrity of its components, (2) the operability and performance of the active components of the systems such as fans, filters, dampers, pumps, and valves and (3) the operability of the systems as a whole and, under conditions as close to design as practical,-the performance of the full operational sequence that brings the 3.1-37 AMENDMENT 8 - OCTOBER 1990 l
Nstemsintocperaticn,includingoperaticnofapplicableportiensofthe protection system, the transfer between normal and emergency power sources, and ;he j
operation of associated systems.
Discussion Due to the non-operating and defueled condition of TMI-2, there are no postulated accidents during PDMS which could result in the generation of fission products, hydrogen, oxygen or other substances which would require containment atmosphere cleanup systems as described in Criterion 41. Therefore, design of the Containment atmosphere cleanup system in accordance with Criterion 43 is not applicable to TMI-2 during PDMS.
3.1.2.40 Criterion 44 - Cooling water.
A system to transfer heat from structures, systems, and components important to safety, to an ultimate heat sink shall be provided. The system safety function shall be to transfer the combined heat load of these structures, systems, and components under normal operation and accident conditions.
Suitable redundancy in components and features, and suitable interconnections, leak detection, and isolation capabilities shall be provided to assure that for onsite electric power system operation (assuming off-site power is not available) and for i
off-site electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.
i Discussion Due to the non-operating and defueled condition of TMI-2, the requirements of Criterion 44 are not applicable.
3.1.2.41 Criterion 45 - Inspection of cooling water system.
The cooling water system shall be designed to permit appropriate periodic inspection-of important components, such as heat exchangers and piping, to assure the-integrity and capability of the system.
Discussion Due to the non-operating and defueled condition of THI-2, the requirements of Criterion 45 are not applicable.
3.1.2.42 Criterion 46 - Testing of cooling water system.
The cooling water system shall ba designed to permit appropriate periodic pressure and functional testing to assure (1) the structural and leaktight integrity of its components, (2) the operability and the performance of the active components of the 3.1-38 AMENDMENT 8 - OCTOBER-1990
t' stem, and (3) the operability of the system as a whole and, under conditiens as y
close to design as practical, the perfomance of the full operational sequence that brings the system into operation for reactor shutdown and for loss-of-coolant accidents, including operation of applicable portions of the protection system and the transfer between nomal and emergency power sources.
Discussion Due to the non-operating and defueled condition of THI-2, the requirements of Criterion 46 are not applicable.
3.1.2.43 Criterion 50 - Containment design basis.
The reactor containment structure, including access openings, penetrations, and the containment heat removal system shall be designed so that the containment structure and its internal compartments can accommodate, without exceeding the design leakage rate and with sufficient margin, the calculated pressure and temperature conditions resulting from any loss-of-coolant accident. This margin shall reflect 1
consideration of (1) the effects of potential energy sources which have not been included in the determination of the peak conditions, such as energy in steam generators and as required by article 50.44 energy from metal-water and other chemical reactions that may result from degradation but not total failure of emergencycorecoolingfunctioning,(2)thelimitedexperienceandexperimental data available for defining accident phenomena and containment responses, and (3) the conservatism of the calculational model and input parameters.
Discussion Due to the non-operating and defueled condition of TMI-2, the requirements of Criterion 50 are not applicable.
However, the intent of Criterion 50 has been addressee by considering that there are functional requirements required to be i
provided by the Containment during the PDMS period which are different, and substantially less than those described in Criterion 50.
The Containment, during PDMS, is required to function as the primary barrier between the radioactive contamination inside the Containment and the environment.
It provides.this function in three ways:
- 1) it minimizes the uncontrolled migration of contamination from inside the Containment to the environment, 2) it functions as an envelope to control the release of Containment atmosphere effluents to the environment, and 3) it functions as primary shielding for the radioactive materials inside the Containment.
The Containment was originally designed and constructed to the criteria described in Criterion 50.
Since the design basis requirements for the Containment during PDMS are substantially less than that required by Criterion 50 and the Containment isolation capabilities were not degraded by either the accident or the recovery activities, the Containment is capable of meeting the requirements of PDMS.-
3.1-39 AMENDMENT 8 - OCTOBER 1990
'3.1.2.44 Criterion 51 - Tracture prevention of containment pressure brundary.
The reactor containment boundary shall be designed with sufficient margin to assure that under operation, maintenance, testing, and postulated accident conditions (1) its ferritic materials behave in a nonbrittle manner and (2) the probability of rapidly propagating fracture is minimized. The design shall reflect consideration of service temperatures and other conditions of the containment boundary material during operation, maintenance, testing, and postulated accident conditions, and the uncertainties in deternining (1) material properties, (2) residual, steady state, and transient stresses, and (3) size of flaws.
Discussion Due to the non-operating and defueled condition of THI-2, the requirements of Criterion 51 are not applicable.
3.1.2.45 Criterion 52 - Capability for containment leakage rate testing.
The reactor containment and other equipment which may be subjected to containment test conditions shall be designed so that periodic integrated leakage rate testing can be conducted at containment design pressure.
Discussion The requirements for leakage rate testing capability as established in Criterion 52 assume the possibility of pressurization events and the consequent potential for leakage of fission products. Due to the non-operating and defueled condition of TMI-2, there are no events which can result in significant pressurization of the Containmentandthreatenitsisolationcapabilities(SeeCriterion50).
During passive pressure control, the Containment will be continually vented via the AFHB to the atmosphere through a passive breather system. Thus, the pressure inside the Containment will be at equilibrium with atmospheric pressure. During operation of the Containment Purge System, a filtered, monitored exhaust path is provided via the station vent to the atmosphere. Therefore, no significant pressurization of the Containment could occur during this mode of operation.
Based on the above conditions, na leak rate te:, ting is required during PDMS.
3.1.2.46 Mierion 53 - Provisions for containment testing and inspection.
The reactor containment shall be designed to permit (1) appropriate periodic inspectionofallimportantareas,suchaspenetrations,(2)anappropriate surveillanceprogram,and(3)periodictestingatcontainmentdesignpressureof the leaktightness of penetrations which have resilient seals and expansion bellows.
Discussion-The Containment has been designed and constructed in accordance with Criterion 53.
Due to the unique condition of THI-2 during PDMS, the specific requirements of Criterion 53 are not applicable; however, the intent of Criterion 53 has been addressed by providing the appropriate surveillance activities based on the Containment isolation requirements for PDMS as described in Section 7.2.1.
3.1 AMENDMENT 8 - OCTOBER 1990
Y.1.2.47 Criterion 54 - Piping systems penetrating containment.
Piping systems penetrating primary reactor containment shall be provided with leak detection, isolation, and containment capabilities having redundancy, reliability, and performance capabilities which reflect the importance to safety of isolating these piping systems.
Such piping systems shall be designed with a capability to test periodically the oper6bility of the isolation valves and associated apparatus and to determine if valve leakage is within acceptable limits.
Discussion Due to the non-operating and defueled condition of THI-2 during PDMS, the specific requirements of Criterion 54 regarding leak detection, isolation, and containment capabilities are not applicable. However, piping systems penetrating Containment have been isolated and will be maintained isolated during PDMS as described in Section 7.2.1.
l 3.1.2.48 Criterion 55 - Reactor coolant pressure boundary penetrating containment.
Each line that is part of the reactor coolant pressure boundary and that penetrates primary reactor containment shall be provided with containment isolation valves as follows, unless it can be demonstrated that the containment isolation provisions for a specific class of lines, such as instrument lines, are acceptable on some other defined basis:
1.
One locked closed isolation valve inside and one locked closed isolation valve outside containment; or l
2.
One automatic isolation valve inside and one locked closej isolation l
valve outside containment; or i
3.
One locked closed isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment; or
~
4.
One automatic isolation valve inside and one automatic isolation' valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment.
Isolation valves outside containment shall be beated as close to containment as practical and upon loss of actuating power, automatic isolation valves shall be designed-to take the position that provides greater safety.
Other appropriate requirements to minimize the probability of consequences.of an accidental rupture of these lines or of lines connected to them shall be provided as necessary to assure adequate safety.
Determination of the appropriateness of these requirements, such as higher quality in design, fabrication, and' testing, 3.1-41 AMENDMENT 8 - OCTOBER 1990
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'a*dditional provisions for inservice inspecticn, protecticn against more severe natural phenomena, and additional isolation valves and containment, shall include consideration of the population density, use characteristics, and physical characteristics of the site environs.
Discussion Due to the non-operating and defueled condition of TMI-2 there is no reactor coolant pressure boundary, therefore, the specific requirements of ".riterion 55 are not applicable. However, the intent of Criterion 55 has been addressed for PDMS.
All piping which penetrates the Containment has been isolated as described in Section 7.2.1.
3.1.2.49 Criterion 56 - Primary containment isolation.
Each line that connects directly the containment atmosphere and penetrates primary reactor containment shall be provided with containment isolation valves as follows, unless it can be demonstrated that the containment isolation provisions for a specific class of lines, such as instrument lines, are acceptable on some other defined basis:
1.
One locked closed isolation valve inside and one locked closed isolation valve outside containment; or 2.
One automatic isolation valve inside and one locked closed isolation valve outside containment; or 3.
One locked closed isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment; or 4.
One automatic isolation valve inside and one automatic isolation valve outside containment.
A simple check valve may not be used as the automatic isolation valve outside containment.
Isolation valves outside containment shall be located as close to containment as practical and upon loss of actuating power, automatic isolation valves shall be designed to take the position that provides greater safety.
Discussion Due to the unique condition of TMI-2 during PDMS, the specific requirements of Criterion 56 are not applicable; however, the intent of Criterion 56 has been addressed,for PDMS.
Piping systems penetrating Containment have been isolated outside Containment and will be maintained isolated.- Due to the non-operating and defueled condition of TMI-2, one closed isolation valve outside Containment on each piping penetration provides suitable Containment isolation during PDMS.
See Section 7.2.1.
3.1-42 AMENDMENT 8 - OCTOBER 1990
f ll */.1.2.50 Criterion 57 - Closed system isolatien valves.
Each line that penetrates primary reactor containment and is either part of the reactor coolant pressure boundary or connected directly to the containment atmosphere shall have at least one containment isolation valve which shall be e
either automatic, or locked closed, or capable of remote manual operation. This valve shall be outside containment and located as close to the containment as practical. A simple check valve may not be used as the automatic isolation valve.
Discussion Due to the unique condition of TMI-2 during PDMS, the specific requirements of Critr ion 57 are not applicable; however, the intent of Criterion 57 has been add:., sed for PDMS.
All piping systems penetrating Containment have been isolated as described in Section 7.2.1.
l 3.1.2.51 Criterion 60 - Control of releases of radioactive materials to the environment.
The nuclear power unit design shall include means to control suitably the release of radioactive materials in gaseous and liquid effluents and to handle radioactive solid wastes produced during normal reactor operation, including anticipated operational occurrences.
Sufficient holdup capacity shall be provided for retention of gaseous and liquid effluents containing radioactive materials, particularly where unfavorable site environmental conditions can be expected to impose unusual operational limitations upon the release of such effluents to the environment.
Discussion Due to the unique condition of TMI-2 during PDMS, the specific requirements for Criterion 60 are not applicable; however, the intent of Criterion 60 has been addressed by providing means to suitably control releases of radioactive materials to the environment during PDMS.
3.1.2.52 Criterion 61 - Fuel storage and handling and radioactivity control.
l l
l The fuel _ storage and handling, radioactive waste, and other systems which may I
contain radioactivity shall be designed to assure adequate safety under normal and
. postulated accident conditions. These systems shall be designed (1) with a l.
capability to permit appropriate periodic inspection and testing of components important to safety, (2) with suitable shielding.for radiation protection, (3) with appropriate containment, confinement, and filtering systems, (4) with a residual heat removal capability having reliability and testability that reflects the L
importance to safety of decay heat and other residual heat removal, and (5) to L
prevent significant reduction in fuel storage coolant inventory under accident conditions.
l 3.1-43 AMENDMENT S - OCTOBER 1990 i
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l.:dscussion Due to the non-operating and defueled condition of TMI-2, the requirements of Criterion 61 with regards to fuel handling and storage are not applicable.
However, small quantities of resiaual fuel remain in various locations within the Reactor Coolant System and in other areas of the Reactor Building; the Defueling Completion Report (DCR) (Reference 3.1-7) identified the quantity of residual fuel in each defined location and addressed the potential for fuel relocation.
As discussed in Section 4.3.4, the criticality analyses provided in the DCR demonstrated that criticality has been precluded at THI-2.
Finally, personnel accessibility, potential exposure, and other protective features for the residual fuel and other radioactive material are provided consistent with the requirements of Criterion 61.
3.1.2.53 Criterion 62 - Prevention of criticality in fuel storage and handling.
Criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations.
Discussion Due to the non-operating and defueled condition of THI-2, the requirements of Criterion 62 are not applicable. See Section 4.3.
.3.1.2.54 Criterion 63 - Monitoring fuel and waste storage.
Appropriate systems shall be provided in fuel storage and radioactive waste systems and associated handling areas (1) to detect conditions that may result in loss of residual heat removal capability and excessive radiation levels and (2) to initiate appropriate safety actions.
Discussion Due to the non-operating and defueled condition of THI-2 there will not'be any raterials which can generate sufficient decay heat to require residual heat removal a.apabilities. Therefore, the requirements of Criterion 63 are not applicable.
3.1.2.55 Criterion 64 - Monitoring radioactivity releases.
Means shall be provided for monitoring the reactor containment atmosphere, spaces containing components for. recirculation of'k of-coolant accident fluids, effluent discharge paths,.and the plant environs for radioactivity that may be released from normal operation, including anticipated operational occurrences, and from postulated accidents.
Discussion Due to the unique condition of THI-2 during PDMS, the specific requirements of Criterion 64 are not applicable; however, the intent of Criterion 64 has been addressed by providing menns to monitor radioactivity releases, as described in 1 Sections 7.2.1.2 and 7.2.4, commensurate with the plant condition during-PDMS.
3.1-44 AMENDMENT 8 - OCTOBER 1990
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o REFERENCES 3.1-1 Letter, Travers, W. D. (NRC) to Standerfer, F. R. (GPUNC),
" Approval of Exemption from 10 CFR 50.61," dated December 30, 1985.
3.1-2 Letter, Snyder, B. J. (NRC) to Kanga, B. K. (GPUNC), "10 CFR 50.49, i
' Environmental Qualification of Electrical Equipment Important to Safety for Nuclear Power Plants'," dated July 22, 1983.
i 3.1-3 Letter, Stolz, J. F. (NRC) to Standerfer, F. R. (GPUNC), " Issuance of Amendment (TAC No. 65337)," dated May 27, 1988.
3.1-4 Letter, Snyder, B. J. (NRC) to Hovey, G. K. (Met-Ed), Re:
Exemption from 10 CFR 50 Appendix.1. dated September 2; 1981.
3.1-5 Letter, Snyder, B. J. (NRC) to Hovey, G. K. (Met-Ed), Re:
Relief from the Inservice Inspection Program Requirements of 10 CFR 50.55a, dated April 27, 1981.
1 3.1-6 GPU Nuclear letter, 4410-90-L-0044, " Decommissioning Financial Assurance Certification Report for... TMI-2," dated July 26, 1990.
3.1 GPU Nuclear letter, 4410-90-L-0012. "Defueling Completion Report, Final Submittal," dated February 22, 1990.
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3.1-45 AMENDMENT 8 - OCTOBER 1990